eGENuclear Lnergy

AdvancedReactorProgramsGeneralElecmcCompany6835ViaDe/OroM/CSanJose,CA95/I9.I315408365.6600

XL-P2A37-94003 January 15, 1994

U.S. Departmentof EnergyDOE San FranciscoOperationsOfficeNuclear Division1301 Clay Street, Room 700NOakland, CA 94612-5203

Attention: Kashmira Mall

Subject: Contract No. DE-AC03-93SF19681"Pu Consumptionin Advanced Light Water Reactors"Transmittal of Phase 1C Report

Enclosedis the GE Nuclear Energy Phase 1C Report "NEDO-32314 - Study of PuConsumptionin Advanced Light Water Reactors; Evaluationof GE's AdvancedBoiling Water Reactor Plants, Compilationof Phase 1C Task Reports'.

If there are any questions, please contact me at (408) 365-6468 or(408) 925-1714.

Edward EhrlichProject ManagerGE ALWR Pu ConsumptionStudy

XL-P2A37-94003 2 January 15, 1994

Attachment: GE Nuclear Energy Report NEDO-32314 - Study of PuConsumptionin Advanced Light Water Reactors; Evaluation ofGE's Advanced BoilingWater Reactor Plants, Compilation ofPhase 1C Task Reports,dated September 15, 1994

cc: Aundra RichardsDOE-SANDave PinesDOE-SANRobertNeuholdDOE-HQ NE-45ErnieCondonDOE-HQ NE-45

NEDO-32314RFP DE-AC03-93SF19681

January15, 1994

@GENuclear Energy

i

SanJose, Ca/ifornia

Study of Pu Consumption inAdvanced Light Water ReactorsEvaluation of GE Advanced Boiling Water Reactor Plants

Compilation of Phase 1C Task Re _orts

NEDO-32314RFP DE-AC03-93-SF19681

JANUARY 1994

TITLE: STUDY OF Pu CONSUMPTION IN LIGHT WATER REACTORSEvaluation of GE Advanced Boiling Water Reactor PlantsCompilation of Phase 1C Task reports

Preparedfor theUnited States Departmentof Energy

UnderContractNo. DE-ACO3-93SF19681

• MAS]ERGENuclearEnergy

AdvancedReactorProgramsSanJose,Ca/ifomia95119-7315

i_TRlt_JTION OF ]'HIS DOCUMENT 1_ _jNL.1Mrl F._L_'_93-426..02

DISCLAIMER

This reportwas preparedas an accountof work sponsoredby an agency of the UnitedStatesGovemment. Neitherthe United StatesGovernmentnor any agencythereof,norany of their employees,norany of theircontractors,subcontractors,or theiremployeesmakesany warranty,expressor implied,or assumesany legal liabilityor responsibilityforthe accuracy, completenessor usefulnessof any information,apparatus, product orprocessdisclosed,or representsthat its use wouldnot infringeprivatelyowned rights.Referencehereinto any specificcommercialproduct,process,or serviceby trade name,trademark, manufacturer, or otherwise, does not necessarily constitute or imply itsendorsement, recommendation,or favoring by the United States Govemment or anyagencythereof.The viewsand opinionsof authorsexpressedhereindo not necessarilystateor reflectthoseof the UnitedStatesGovernmentorany agencythereof.

Study of Pu Consumption in Advanced Light Water Reactors

Evaluation of GE-ABWR

TABLE OF CONTENTS FOR PHASE 1C REPORT

Phase IC WBS

SUMMARY REPORT OF PHASE 1C EVALUATIONS

1.0 CORE AND SYSTEM PERFORMANCE

1.1 Reference Spent Fuel Design 1.1-11.1.1 Normal Operation 2.1 1.1.1-11.1.2 Transient Response of Reference Fuel Design 2.2 1.1.2-11.1.3 Fuel Characteristics after Irradiation 1.2 1.1.3-1

1.2 Alternate Core Designs for Pu Disposition 2.1 1.2-11.2.1 Alternatives for 100 MT in 25 Years 2.1 1.2-21.2.2 Altematives for 50 - 100 MT in 25 Years 2.1 1.2-51.2.3 Alternatives for 100 MT in more than 25 Years 2.1 1.2-7

1.3 Relationship between Pu Enrichment, Discharge Exposure, 1.3 1.3-1Disposition Time, Isotopics and Number of Reactors

2.0 FUEL CYCLE

2.1 MOX Fuel Fabrication Requirements for Various Spent 1.7 2.1-1Fuel Scenarios

2.2 MOX Fuel Handling and Disposal2.2.1 Criticality Analyses for Storage, Handling & 1.3 2.2.1-1

Repository2.2.2 Spent Fuel Disposition in Repository 1.4 2.2.2-12.2.3 Spent Fuel Proliferation Resistance 1.4 2.2.3-

2.3 Qualifying and Licensing MOX Fuel2.3.1 Review of MOX Fuel Licensability 1.6 2.3-12.3.2 Program Plan for Lead Fuel Testing 1.6 2.3-112.3.3 US Infrastructure for Lead Fuel Testing 1.5 2.3-202.3.4 European Infrastructure for MOX Testing 4.6 2.3.4-1

Table of Contents for Phase 1C Report (Continued)

Phase IC WBS

2.0 (Continued)

2.4 MOX Fuel FabricationFacility Requirements 1.72.4.1 Process Simulation 1.7 2.4.1-12.4.2 Weapons Pu Interface: Input Pu Specifications 1.1 2.4.2-12.4.3 Layout, Cost, Schedule, Rate of Pu Processing 1.7 2.4.3-12.4.4 First-of-a-Kind Technologies 1.8 2.4.4-1

2.5 Waste Stream Characterization/Management 1.9 2.5-1

3.0 TRITIUM PRODUCTION

3.1 MOX Core Design for Tritium Production 2.1 3.1-1

3.2 Tritium Target Design and Performance 3.4 3.2-1

3.3 Tritium Target Fabrication and Recovery Facility Requirements 3.1 3.3-1

3.4 ABWR Plant Operations for Tritium Production 3.3 3.4-1

4.0 INFRASTRUCTURE AND DEPLOYMENT

4.1 Non-U.S. Facilities Technology Evaluation4.1.1 Japanese MOX Fabrication Facilities 4.6 4.1.1-14.1.2 BNFL Facilities and Experience 4.6 4.1.2-14.1.3 Comparison of U.S. and Foreign (UK) MOX Fuel 4.6 4.1.3-1

Fabrication Facility Regulatory Requirements

4.2 Adapting Commercial MOX Fuel Fabrication Experience 4.6 4.2-1

4.3 Pu Disposition Complex Infrastructure 4.1,4.2 4.3-1

4.4 Transportation Infrastructure4.4.1 Transport Logistics for Tritium Production 3.1 4.4.1-14.4.2 Transportation of Plutonium Materials for MOX 1.10,4.3 4.4.2-1

Fabrication Facility4.4.3 Transportation of Nuclear Waste 4.4 4.4.3-14.4.4 Spent Fuel Transportation and Logistics 4.5 4.4.4-14.4.5 Comparison of U.S. and International Transport 1.10 4.4.5-1

Regulations

Table of Contents for Phase 1C Report (Continued)

Phase IC WBS

5.0 SAFETY AND ENVIRONMENTAL APPROVAL

5.1 Pu Disposition Complex Safety Approval with 6.2 5.1-1Tritium Production

5.2 Impact of Tritium Production on Environmental 6.2 5.2-1Approval

5.3 ABWR Disposition Complex Safety Approval 6.2 5.3-1Program

5.4 Environmental Permitting Plan and Schedule 6.2 5.4-1

6.0 DEPLOYMENT REQUIREMENTS

6.1 Development Requirements Overview 5.1 6.1-16.2 Development Requirements for MOX Factory 5.2 6.2-16.3 Development Requirements for Tritium Production 3.2,5.3 6.3-1

7.0 SAFEGUARDS AND SECURITY

7.1 Safeguards Requirements for Pu Transport 1.10 7.1.1-1

8.0 COST AND SCHEDULE

8.1 Cost and Schedule Analysis 6.1 8.1-1

Appendix A: Compliance of MOX Fueled GE9 Assembly with Amendment 22 of A-1NEDE-24011-P-A (GESTAR II)

Appendix B: Repository Considerations B-1

Appendix C: T2P2: A Computer Program for Estimating Tritium Target Performance C-1and Tritium Environmental Source Terms

Appendix D: Radiological Safety Requirements and Criteria for the Sellafield MOX D-1Plant

SUMMARY REPORT OF PHASE 1C EVALUATIONS

Contract No. DE-ACO3-93SFI9681, "Pu Consumption in Advanced Light Water Reactors"

The evaluations conducted during Phase 1C of the Pu Disposition Study have provided further

results which reinforce the conclusions reached during Phase 1A & 1B:

• 3E's ABWR was designed for a full core loading of MOX fuel and requires no reactormodifications or plant systems level changes to use MOX fuel.

• The ABWR design allows a wide flexibility in full MOX core design options to meeta wide range of disposition objectives.

• The technology for converting weapons Pu to MOX fuel has already beendemonstrated by DOE complex activities and by commercial operations.

• Existing DOE facilities can be adapted to building the MOX factory.

• No fundamental technical issues relative to nuclear safety, worker safety, publichealth and environmental impact or licensing have been identified which wouldpotentially delay either a MOX plant or reactor construction or startup.

• Institutional organizations and criteria which are needed to implement the MOX plantcan be established to support the project schedule.

• The infrastructure exists for near term (8-10 years) deployment in the U.S. of theABWR Plutonium Disposition Complex.

These conclusions clearly establish the benefits of the fission option and the use of the ABWR

as a reliable, proven, well-defined and cost-effective means available to disposition the weapons

Pu. This project could be implemented in the near-term at a cost and on a schedule being

validated by reactor plants currently under construction in Japan and by cost and schedule

history and validated plans for MOX plants in Europe.

Evaluations conducted during this phase have established that (1) the MOX fuel is licensable

based on existing criteria for new fuel with limited lead fuel rod testingl (2) that the applicable

requirements for transport, handling and repository storage can be met, and (3) that all the

applicable safeguards criteria can be met.

During this phase, visits were made to DOE's Complex 21 sites to assess the existing

infrastructure that might support the disposition process. Contact was established with LLNL

and LANL to determine the technical capabilities and interfaces that might be implemented at the

front-end of the MOX factory. Transportation requirements and infrastructure for all aspects of

the disposition process from fresh fuel to spent fuel were defined and evaluated. The

infrastructure for carrying out this disposition - management and staff, facilities, and procedures -all exist.

Evaluation of the capability to produce contract quantities of tritium was demonstrated in Phase

1A using a conventional urania fueled core. During this phase, a MOX core design was

developed and analyzed which shows that the same requirements can be safely, met with a MOX

fueled core such that tritium production and Pu disposition can be carried out concurrently.

Cost and schedule evaluatiops are ongoing and additional data that have been collected are

consistent with previously reported cost and schedule estimates.

The individual task results are summarized below.

CORE AND SYSTEM PERFORMANCE

A nominal candidate spent fuel MOX core design has an average Pu enrichment of 3.5% and an

average discharge exposure level of 37,000 MWD/MT(typical of the upper range of discharge

exposure for GE current 8x8 ABWR fuel design). Neutronics and safety analysis show large

margins to safety limits with no reactor system changes and no core uncovering under

accident conditions. For the baseline 25 year project term case, this design requires up to 6

reactors for dispositioning 100MT of Pu. For a disposition time of 60 years (the design lifetime

of the ABWR), only 2 reactors are required. For a disposition campaign of 40 years - the current

license term without relicensing - either the disposition amount could be lowered to 75 MT or a

third reactor added for a full 100 MT campaign.

With t,_e flexible capabilities of the ABWR, alternate core design options are available which

permit the plutonium to be dispositioned using fewer reactors. Options include discharging the

fuel at a slightly lower exposure or increasing the plutonium enrichment. It is possible to

disposition 100 MT of plutonium with two reactors in 36 years in a core design with 5%

enrichment and 37,000 MWD/MT burnup. Another core design option requires only one reactor

in a 54 year campaign to disposition the same amount. This option is a core design of 5%

enrichment and 30,000 MWD/MT burnup which produces discharge isotopics and bundle

radiation levels comparable to the current average BWR discharge exposure.

Explicit relationships between these variables, in particular the effect of disposition time, can be

seen in the figure below and are discussed further in the report. The cost tradeoffs for these

options are continuing. It should be noted that generically, for higher enrichment LWR designs,

licensing delays and technical development risks are increased.

100 Mt of Pu at 40 GWDIMt

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FUEL CYCLE

MOX Throughput Requirements

The mixed oxide fuel fabrication throughput requirements have been established for four spent

fuel scenarios for a 3.5% Pu enriched MOX fuel that is discharged at 38,000 MWD/MT. The

four scenarios with the resulting fabrication requirements are shown on the accompanying table:

Soent Fuel Scenari9,

Characteristic 1 2 3 4

Available Plutonium, (MT) 100 100 50 50

Time to Disposition, (Yrs.) 25 60* 25 60*

ABWRs required, (Number) 6 2 3 1

Fabrication Plant Design:

MOX Throughput, (MT/Yr.) 220 80 110 40

*ABWR design lifetime

These throughput requirements include a 15% excess production capability margin. The MOX

throughput requirements for the alternate core designs using 5% Pu enrichment are currently

being evaluated.

MOX Fuel Licens#Lg Evahtation and Lead Fuel Testing

GE's new fuel designs are licensed using a process agreed upon between GE and NRC which

requires GE to examine specific documented criteria in detail. The criteria of this document were

considered in detail and indicate that the available data base is sufficient for a preliminary license.

This evaluation included a consideration of all salient MOX fuel properties, in particular as to

how they differ from the properties of urania fuel and the impact of this difference on fuel

performance and licensing criteria. The sole remaining issue concerns the verification and

qualification of a fabrication process to demonstrate that MOX fuel with an acceptable

microstracture can be made and that it performs as expected in in-reactor experiments. No need

for full scale assembly tests was identified, in particular because prototypical fuel designs are

unlikely to be compatible with a predominantly urania core and also because such isolated

assemblies do not provide system level response information. Nevertheless, if there is a need to

irradiate full scale MOX assemblies, the infrastructure for doing such testing is readily available.

A short full MOX core confirmatory test activity is already reflected in the Pu Disposition

Project schedule.

MQX Fuel Hcmdlmg and Disposal

The work scope elements during Phase 1C called for an examination of the spent nuclear fuel

(SNF) characteristics relative to proliferation resistance, handling, storage and repository

requirements. Although the specific requirements are yet to be fully enunciated for some of these

categories, the evaluations show that the requirements which exist at this time can be met. In

particular, it was found that the proliferation resistance as measured by the attractiveness level of

the weapons plutonium is degraded with each process step even prior to irradiation as fuel, and

that the fission option proposed here has better and proven proliferation resistance compared to

other non-fission disposition options. For handling, storage and disposal, all applicaLle criteria, in

particular subcriticality requirements, are met while employing existing technology and

configurations used for commercial SNF.

MOX Fuel Fabrication Facility Reqmrements

A systems analysis for establishing the requirements of a mixed oxide fuel fabrication plant that

will provide the capability needed for the plutonium disposition mission vvas carried out. The

analysis approach includes ,_ dynamic simulation of the entire disposition process from the

accumulation of excess plutonium metal to the final disposal of the spent fuel in long term

storage. All system interfacing requirements among the storage facility, plutonium shipping, fuel

fabrication, fresh fuel shipping, reactor operation, spent fuel shipping and final long term storage

can be evaluated. A detailed process simulation model of the fuel fabrication plant was

developed to establish process performance requirements. The dynamic analysis was planned to

provide insight on the optimum strategies for fuel fabrication and to minimize life-cycle waste

accumulation. This dynamic model is currently being evaluated for consistency of numerical

results. Further model development could include elements for establishing the requirements for

real-time material accounting and the requirements for minimizing radiation exposure of workers.

This system analysis approach and dynamic modeling could be implemented in the project phase

to optimize the details of the various activities.

TRITIUM PRODUCTION

The core design for tritium production using conventional urania fuel was presented in the Phase

1A report. During this phase of the study, a MOX fueled core has been designed that meets the

tritium production requirements. This core has a core-average Pu enrichment of 5.9%, uses four

tritium target rods per assembly as before, and utilizes the MOX fuel to an exposure of 28,000

MWD/MT. All the nuclear and thermo-mechanical design criteria for normal operation have been

met. To meet a requirement to design within the already existing target rod database, it would be

necessary to discharge all the target rods every year. An alternative would be to extend the

irradiation data base which would allow the target goals to be met with a variety of fuel cycles of

longer duration which would be more compatible with commercially attractive electricity

production fuel cycles of 18-months or partial core reloads with more frequent refueling.

IN RASTRUCFURE AND DEPLOYMENT

Planned MOX Fuel Fabrication Facifities in Foreitm Ctnmtries

As part of the infrastructure evaluations, the MOX fuel fabrication capability in Japan and

England was considered with a view to technology assessment. The Japanese MOX program is

still in a planning stage. It calls for reprocessing of LWR spent fuel and production of about 100

MT of MOX fuel per year by the year 2002. A reprocessing facility is under construction in

Rokkasho Mura. The MOX fuel fabrication factory is currently in the,design phase. The

planned MOX facility will be fully automated, and is also being designed to accommodate non-

remote maintenance, if this is required. The input feed material to this plant is planned to be a

50-50 master blend of urania and plutonia powder. Transportation casks for international

shipments have been designed and fabricated and are in use. Casks for local shipping are

currently being designed. It is believed that safeguards will be implemented primarily through

IAEA inspection and standards.

A detailed description of the planned MOX fuel facility of BNFL at Sellafield which will use

plutonium from reprocessed commercial SNF is provided in this report. In addition to the

facility descriptions which indicate that the technology is ready and is easily adapted to

dispositioning weapons plutonium, a comparative evaluation of the licensing and safety

requirements for the Sellafield plant vs. a Greenfield facility in the U.S. has also been provided.

Ada_vtit_ Exist#Tg MQX Fuel Fabrication Technology

Although MOX fuel has not been fabricated in significant quantities in the U.S. in nearly two

decades, other countries are proceeding to implement plans for fabrication of MOX fuel from

reprocessed plutonium. Since this technology exists and is being implemented, the changes

needed to adapt this technology for processing weapons plutonium were defined. In most of the

areas such as shielding, worker exposure, handling, and maintainability, weapons plutonium

would be easier to process than reprocessed plutonium. Two areas were identified which require

additional evaluations in adapting this already available technology: the first concerns a re-

evaluation of the constraints - in the form of allowable qaantities of plutonium in any given

area/container- arising from criticality requirements and its effect on throughput, and the second,

surveillance/accountability instrumentation based on gamma activity which may need

modification or alternate instrumentation, as the weapons plutonium g activity is far lower than

in the case of reprocessed plutonium.

Plutonium Disposition Infrastructure in the United States

It was concluded in Phase 1A that the infrastructure appeared to be established for deployment

of an ABWR Pu Disposition Complex in the United States. The brief surveys of each of the

DOE sites conducted under Phase 1C support this conclusion.

The Department of Energy (DOE) already has thesites and capabilities, and the flexibility with

these sites and capabilities, to deploy an electric power producing, full MOX-fueled ABWR Pu

Disposition Complex. For study comparison purposes, the reference case for deployment of the

Plutonium Disposition Complex in the United States is a new "Greenfield," in which facilities are

constructed and located all together on a hypothetical site at Kenosha, Wisconsin. The purpose

of the infrastructure portion of this study was to determine the extent to which the "complex"

could utilize the existing capabilities at one or more of the existing DOE and/or commercial sites.

The study included visits and a collection of data for the following sites:

• Idaho National Engineering Laboratory (INEL)• Nevada Test Site

• Oak Ridge Reservation (ORR)• Pantex Plant

• Savannah River Site (SRS)° Hanford Site

• Lawrence Livermore National Laboratory (LLNL)° Los Alamos National Laboratory (LANL)

It is clear that considerable cost effective, installed capability is available within the DOE

community now for meeting the Pu disposition needs in the near term with one or more electric

power producing, full MOX-fueled ABWR plants. These capabilities can be implemented in the

short term with effort ranging from minor refurbishing to upgrading of existing facilities, with

only a few requirements, such as the reactor, being Greenfield efforts at all sites. It is anticipated

that a minimum cost deployment will be to locate the entire Pu Disposition Complex at one site.

SRS, ORR and INEL already have in place significant applicable elements. It is also possible to

take advantage of unique capabilities which exist at these and other sites and create a distributed

"complex," with some additional cost for transportation between sites.

WASTE CHARACTERIZATION AND TREATMENT

Waste stream characteristics were updated where new information modified previous estimates.

Treatment and disposal options were also identified for each waste stream. It was concluded that

existing and planned treatment and disposal technology will be adequate, and that the types and

quantities of wastes generated are typical of normal reactor operations.

SAFETY AND ENVIRONMENTAL APPROVAL

The safety approvals and environmental permitting activities required for the ABWR Plutonium

Disposition Complex were examined. The New Production Reactor (NPR) safety approval was

used as a model for the development of a detailed ABWR Pu Disposition Program safety

approval schedule. This schedule assumed that a single Integrated Safety Analysis Report

(ISAR) will be submitted in stages to DOE and supporting government agencies for approval to

support critical program decision points. The environmental permitting process required for a

Record of Decision (ROD) was also examined for both a Greenfield site and an existing DOE site.

SAFEGUARDS AND SECURITY

Work continued on assessing the impacts of safeguards and security requirements for plutonium,

tritium and enriched lithium on the configuration and operation of the ABWR Plutonium

Disposition Complex. The results of these additional studies indicate that because the number of

shipments is relatively small ( -- 1 fresh fuel shipment/month for the 2 reactor case and less for

enriched lithium or tritium), transportation of controlled materials is unlikely to be a controlling

factor in the configuration of the complex. However, as discussed in Phase 1A, maintaining the

uncertainty in material accountability at acceptable levels must be addressed given the large

quantity of plutonium to be processed.

COST AND SCHEDULE

Assuming a national commitment, and the application and use of existing ABWR submittals,

existing DOE site-specific environmental data and procedures, and the use of existing DOE and

MOX fuel fabrication technology, it is concluded that the overall ABWR Pu Disposition

Complex schedule issued in the ABWR Phase 1A report can be achieved. Further refinement of

the ABWR Plutonium Disposition Complex costs and schedules was initiated. The structures

and improvements account (EEDB account 21) of the baseline ABWR capital cost estimate was

reviewed against current information available from construction of the two ABWRs in progress

in Japan and similar cost studies done for the GE Simplified Boiling Water Reactor (SBWR).

New cost estimates, cash flows and schedules are being developed for disposition of 50 or 100

MT of plutonium in 25 years, 40 years and for the reactor lifetime of 60 years. Development of

preliminary revenue calculations and review of cost and schedule data from an existing European

MOX fuel fabrication facility owner/operator was begun. Verification of environmental and

safety approval schedules was completed. Evaluation of the cost tradeoff from the use of

existing facilities at DOE sites vs a Greenfield site was initiated.

1. CORE AND SYSTEM PERFORMANCE

GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize full

core loading of Mixed Uranium-Plutonium Oxide (MOX) fuel. This design characteristic and

large power rating allows the ABWR a broad array of core design options to optimally dispose

of weapons plutonium. Whether the objective is to maximize the return on investment or to

minimize initial capital cost, the ABWR with its well established cost and schedule data base,

provides the best means of LWR fission based disposition.

Phase 1A of this study, concluded in May 1993, presented three disposition alternatives.

Subsequent DOE reviews have indicated that the spent fuel option presented the most optimal

means of disposition. Core design efforts during Phase 1C have been focused on this option.

Specifically, the following Agreements and Commitments were issued:

(a) GE to focus on core design options that will result in the fuel being ,lischarged as

spent fuel with isotopic compositions typical of current BWR commercial spent

fuel.

(b) GE to evaluate disposition options where the reference amount of Pu to be

dispositioned will remain at 100MT and the disposition time is limited to 25

years, but information may also be provided for a range of amounts from 50 MTto 100 MT.

(c) GE to evaluate disposition options for the reference amount of 100 MT of Pu and

disposition time of 25 years, but information may also be provided for mission

duration up to the design lifetimes of the reactors. Emphasis will be on 40 year

design life, but 60 year case can also be considered.

The first subsection following describes the Reference Spent Fuel Case. Both neutronic and

transient results are given for this case. The next section shows various spent fuel alternatives

for differing dispositon goals. The final section has two parts. First, a generic discussion is

presented on the relationship between the quantity of Pu that has to be disposed, the disposition

time, Pu enrichment and exposure. Following this, the relationship between spent fuel isotopics,

Pu enrichment and exposure is examined.

1.1.0-1

1.1 REFERENCE SPENT FUEL DESIGN

The general approachutilized in the core and fuel nuclear design for the Reference Plutonium

Spent Fuel option was to establish the bundle average enrichment to enable the fuel to attain the

highest allowable batch average discharge exposure consistent with the thermal-mechanical

limits imposed on the GE9 fuel product line which is approximately 38,000 MWd/metric tons.

The batch size utilized in this design is 232 bundles in equilibrium (a batch fraction of 26

percent) which results in a batch average discharge exposure of 37.0 GWd/metric tons. The

resulting maximum residence time for these bundles is, assuming the reference capacity factor of

75 percent, six years. This discharge exposure is consistent with the upper range normally

associated with urania spent BWR fuel.

The reference plutonium spent fuel designpresented here can disposition the required plutonium

inventory in just over nineteen years operation of six ABWR power plants. A summary of the

important fuel cycle parameters and plutonium consumption rates is shown in Table 1.1-1.

Table 1.1.1. Key Parameter Summary, Pu Spent Fuel Case

Numberof reactors 6Cycle Length,EFPD 392.2DischargeExposure,MWd/metric tons 37081ReloadBatch Size, bundles 232Plantcapacityfactor 75%Pu Loadingkg/bundle 5.33Pu Consumption rate metric tons/yr 5.19MOX fuel rod usage, rods/year 58000

1.1-1

1.1.1 NORMAL OPERATION

The important features of the core design for the reference Plutonium Spent Fuel Option are

summarized in the following sections. The reference fuel bundle and core design are described

in the following sections. A description of operating limits follows, showing thermal margins,

reactivity margins, reactivity coefficients and the operation of the core through an equilibrium

cycle.

1.1.1.1 Reference Bundle Design

The bundle design for the spent fuel option contains plutonium in all sixty power producing

rods. This design also has a relatively large number of gadolinia bearing burnable poison rods.

This bundle design closely resembles the reactivity characteristics normally associated withenriched uranium fuel.

The bundle axial and radial enrichment distribution is given in Figure 1.1.1-1 along with the

gadolinia and plutonia concentration distribution. Values of enrichment for uranium are read in

hundreds of a percent. For instance, the "071" refers to 0.71 w/o U:35. Only natural uranium is

used. Two reasons for this are that a slight contribution from the U235aids the reactivity

coefficients and natural uranium is thought to be easier to use in the fuel fabrication facility,

since quality control on the material is slightly higher for natural than for tails. The plutonia

concentrations (of isotopic composition given previously) are similarly given in hundreds of

weight percent. The minimum plutonia concentration is 1.00 w/o and the maximum is 4.20 w/o

PuO2. Use is made of only one gadolinia concentration (1 w/o). There are a total of seven pellet

types required for this option, three of which are gadolinia rods.

The infinite lattice radial powe_ peaking is essential in determining the peak power producing

rod. The distribution of relative power peaking (normalized to unity across the lattice) is shown

at beginning of life for the 40 percent void case in Figure 1.1.1-2. The maximum infinite lattice

relative power peaking factor is shown as a function of exposure for the 0 percent, 40 percent

and 70 percent void history cases in Figure 1.1.1-3.

1.1.1-1

Figure 1.1.1-1 GE9 Bundle Design For the Reference Pu Spent Fuel Alternative

i ii i i

iiiiii iii

• - . 3 }: : :: - : Bundle Average.w v v _.. U23sEnrichment = 0.69 w/o

_ Q __ _ __ _ _ @ Total IN/Loading = 5.33 gg

Q Q Q _ _ O _ (_ Average Fissile Fraction = 3.48 w/oMass of Heavy Metal = 179.0 Kg

_______ ActiveFuel Length = 381.0cm

®®®®®O®®®®®®®@®®®®®®®®®Q

Enrichment> 071 071 071 071 071 071 071

Plutonia > 100 160 230 28{ 280 320 420

Oadolinia > 100 100 100

I.I.I-2

Figure 1.1.1-2 Beginning of Life Lattice Power Distribution For the

Reference Pu Spent Fuel Alternative

1.302 1.273 1.287 1.308 1.309 1.288 1.274 1.305

i iii iiiii

1.273 0.788 0.706 0.760 0.761 0.706 0.789 1.276ii

1.287 0.706 0.697 0.763 0.763 0.697 0.707 1.290

"i.... .'._i!::

1.308 0.760 0.763 _i_! 0.763 0.761 1.311 i!il _,

1.309 0.761 0.763 _!_i_!i_::0.764 0.761 1.312

1.288 0.706 0.697 0.763 0.764 0.697 0.707 1.291i

1.288 0.789 0.707 0.761 0.761 0.707 0.789 1.277ii

1.305 1.276 1.290 1.311 1.312 1.291 1.277 1.307

I.I.I-3

Figure 1.1.1-3 Lattice Power Peaking

11 i iiiiii i i iii i I

1.35!_ .....

_0_ Voidlao - _ 40% Void

oooo 70% Void

•_ !.20 ......

1.15 ......•

%qb%%

% qb qt, % qb

1.05 _._,,,," . - .

I'000 5 !0 15 20 25 30 35-- 'dlO 45 50 55

. LatticeAverageExposure(GWD/ST)Figure 1.1.1-4 Hot Uncontrolled K-infinity

1.20,.1 /--. '.....1.15_ // _.,__%\\ .... 0 % Void

J / / -_',_, " ' 40% Void

110_ //.." " " .__, "'" 70%Void

10.90 "-..0.85

, ...... ,, ,.,, '--_ '-_O.8j.

0 10 20 30 40 50LatticeAverageExposure(GWD/ST_

1.1.1-4

The exposure-dependent k. are given in Figure 1.1.1-4 for the uncontrolled lattice. This figure

shows the k.. for three void histories. As seen from Figure 1.1.1-4, the design of the lattice is

such as to emulate a UO, lattice. This is accomplished by means of the widely dispersed

gadolinia in the interior rods.

1.1.1.2 Equilibrium Core Design

The equilibrium core design philosophy was to simulate the reactivity distribution of an

equilibrium UO_ core in order provide simple operation. Also, the fuel fissile inventory was

matched to the residence time of each batch in order to minimize any discharged plutonium

inventory. As stated previously, the target discharge exposure is the maximum allowed by the

thermal mechanical limits on the fuel. A single fuel nuclear design was utilized in an

equilibrium batch of 232 bundles. The detailed core design layout is presented in Figure 1.1.1-5.

The numbers shown in the beginning-of-equilibrium-cycle core map represent the relative

number of the cycle since fuel loading. For instance, the number ',1" refers to fresh fuel (loaded

this cycle) and the number "4" refers to bundles which are about to start their fourth cycle.

A single nuclear design of fuel is loaded into the equilibrium cycle. The important fuel bundle

parameters were summarized previously. A control cell core loading strategy that contains 37

control cells was utilized. Due to the improved hot to cold reactivity swing characteristics of the

ABWR core, it was possible to design the fuel with a large cold shutdown margin and still

maintain sufficient hot excess reactivity. The hot excess reactivity dictated the use of 37 controlcells.

The important parameters of the equilibrium cycle design are summarized in Table 1.1.1-1.

Examination of the results reveals that all thermal, reactivity and energy requirements aresatisfied.

1.1.1-5

Figure 1.1.1-5 Equilibrium Cycle Loading Pattern

_ iii iiiiii il ii ii i i iiii iii ii

4,

1.1.1-6

Table 1.1.1.1. Equilibrium Cycle Key Parameter Summary,

Plutonium Spent Fuel Case

Cycle. Length, EFPD .... 392Cycle Energy, GWd 1,539,,,, ,,,,,

Cycle Exposure, MWd/metric tons 9861

CoreMass, metric, tons. 149.1Reload Enrichment, W/9 U-235 3.70Reloa_dBatch.Siz e, bundles .......... 232Maximum MAPRAT 0.88

-- ,,, -- ,,,,,

Maximum CPRRAT 0.77-- ,, ,=, =,,, ,, , ,

Maxi_mum LHGR, KW/ft ( LHGR limit = !.4.4 ) 13.0MCPR_( OLMCPR = 1.25 ) 1.63Minimum Cold Slautdown Margin "' 1.9,,=, - ,,

Hot Excess Reactivity at BOC ........ 0.5

1.1.1.3 Core Thermal Margins

The critical power ratio and MAPLHGR thermal margin performance are plotted as a function

of cycle exposure in Figure 1.1.1-6 and Figure 1.1.1-7. Operation within MAPLHGR limit

assures the mechanical integrity of the fuel rods is maintained by limiting their power output in

an appropriate manner throughout their lifetime. These results demonstrate ample margin to

core thermal limits.

1.1.1-7

Figure 1.1.1-6 Equilibrium Cycle Maximum MAPRAT vs Exposure

• 1.10

1.05

1.IX) .i i

0.95

0.90

o,o__" ----0.75

0.70

0.65

0.60 1 2 3 4 5 i_ 7 II . 10

C_ck_, GWD_

Figure 1.1.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure

1.10

1.05

1.00 i

0.95 ....

0.90 _

_T 0.95

0.00 ....+., ,..._

o__ _- -o.7o----______0.95 ....

Oii6 0 S ipO 240 _00 4e 0 6. 0 600 7il 0 840 .... 9dlO 10'0

1.1.1-8

1.1.1.4 Reactivity Limit Summary

The reactivity performance of the plutonium spent fuel option design is summarized in Figure

1.1.1-8 and Figure 1.1.1-9. Due to the improved hot to cold reactivity swing of the ABWR N-

lattice, there is abundant cold shutdown margin; therefore, there is little or no impact of the

mixed-oxide fuel utilization on core design from cold shutdown margin considerations.

1.1.1.5 Reactivity Coefficients

The core dynamic void coefficient and Doppler coefficients of reactivity are plotted as a

function of cycle exposure in Figure 1.1.1-10 and Figure 1.1.1-11. The dynamic void

coefficient is somewhat larger than the upper generic limit for ABWR while the Doppler

coefficient is within the generic limits. The significance of these results is discussed in detail in

Section 1.1.3.

1.1.1.6 Core Performance Description

The core performance characteristics as a function of exposure through the cycle are given in

Figure 1.1.1-12 through Figure 1.1.1-17. The core maps in these figures show the control blade

patterns in the core expressed in terms of notches (which are 3-inch sections of blade) withdrawn

from the top of the core. Those cells which have no numbers represent cells in which there are

no blades inserted. The thermal limits and reactivity margins associated with the given exposure

are noted in the summary included with each figure. As seen from these figures, all thermal and

reactivity margins are met. The resulting core average power and exposure profile are also

given. Typical of the ABWR, the power profile shifts towards the top of the core during the last

quarter of the cycle. Since the reactors core design itself provides sufficient margins, it is not

necessary to axially grade the fuel assembly to accommodate the shift in power.

1.1.1-9

Figure 1.1.1-8 Equilibrium Cycle Hot Excess Reactivity

2.00

1.75

i

1.00

% AK ,_0.75

0.50 .... _X

0.2S _,Oo 1.o _o a_ 4.0 s.o 0.o 7.0 e.o o.o lo.o

Cyck Exposure. OWD/st

Figure 1.1.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin

&0

6.6

6.0 ------- _ _ _..___ _

4.0 ........

3.S_hu°Id

tdown a.0M_usm

_5i

P-0

I.S Design1.0 Limit

1.00.5

00 1.0 2.0 S.0 4.0 5.0 e.0 7.0 8.0 9.0 lU.0Cyck_Sxlmure,OWD/_

l.l.l-lO

Figure 1.1.1-10 Equilibrium Cycle Dynamic Void CoefficientI i

0

="1

='2i I Ili I I I I I

='3

-4

m 5 ii , ,

m 6 =--

Void -7

Coef -8

('I_) -o , ,,-10 _ ,,,

-11

_,, .-.-:_.,_------ -- ___-13 ___-14

-1So 1.o 2.o 3.o 4.0 s.o e.o 7.0 e.o o.o lo.oCycleExixm_,GWO/st

Figure 1.1.1-11 Equilibrium Cycle Doppler Coefficient

l.l.l-ll

Figure 1.1.1-12 Equilibrium Cycle Core Data Summary at 0 MWD/stI I I II l lI I III I

Cycle Exposure,MWD/st 0.0

Cycle Energy,MWD 0.0

Numberof Full I_wer Days 0.0I III I

AverageVoidFraction 0.4214IIIIIIII

2 Con: Row, Mlb/hr ' 1.151

MaximumChannelPeaking 1.34762 2 CoreAxial PowerPeak 1.1928

I I

I RAPLHGR 0.7830I II IIII

2 2 MaximumCPRRAT 0.7374

HotExcessReactivity,% 0.83

2 2 Cold ShuRIownMargin 5.18H IIIII I

I

2

[_N - Numberof 3 inchincrementsthatthecontrolbladeis withdrawnfromfullyinserted

26 26

20 , \l..... 20 _

I0 _ I0

| II --

4 ,, 4

2 J 2

°o 0.2 0.4 0.6 0., _.0 s_ _.4 s.6 s., 7,.0 °o 2 4 6 s _0 n 14 _6 n2022 u 26 2, 3o

Core Axial AverageRelative Power Core Axial AverageExposure,GWD/st

1.1.1-12

Figure 1.1.1-13 Equilibrium Cycle Core Data Summary at 2000 MWD/st

I t

I I

CycleExposure,MWD/st 2000.0

CycleEnergy,MWD 344082tt ii it

Number of Full POwerDays 87.6tltt ttt t

CmeAverageVoidFraction 0.4301i II i I

6 6 Core Flow, Mlbjhr 1.151" ' 1.30 5MaximumChannelPeakingt

6 4 6 Core AxialPowerPeak 1.2005

MaximumRAJPLHGR 0.8272

4 4 I MaximumC'PRRAT 0.7119

[ Hot F.x.ceu Rmctivity. % N/A

6 4[ 6 Cold ShuldownMargin N/A

6 6

[_N Numberof 3 inchincrementsthatthecontrolbladeis withdrawnfromfully insertedi

26 26,

' I24 _ _ 24 _t_

• I22 22

18 18

16 16

14 14

Z n 12

1o IO8 II

2 .A 2 "_'rl!

00 0.2 0.4 0.6 0.l 1.0 1.2 IA 1.6 1.8 2,.0 00:2 4 6 8 I0 12 14 16 II 20 22 24 26 21130

Core Axial AverageRelativePower CoreAxialAverageExposure,GWD/st

I.I.I-13

Figure 1.1.1-14 Equilibrium Cycle Core Data Summary at 4000 MWD/st

iiiii i i i ii I ii i i ii i ii i iii i illl

CycleEnergy,MWD 688164iiiii Hill I i

28 Numberof Full PowerDays , 175.3i l I ii I iiii i I

- CoreAverage VoidFraction 0.434 Ii I i ii i iii iiii

0 0 0 Core Flow, Mlb/hr 1.151..... u_Chnel ......Maxim Peaking 1.3581

28 0_ 0 CoreAxial Power Peak 1.2133"RAPLHGR....Maximum 0.8725

" ' ' ' CPRRAT ' 'o o o Maximum 0.7166, , ,,,,,|,,

Hot ExcessReactivity,q6 1.46_I , ' ,,,, , !

28 ] 0J ' 0 28 Cold ShuMownMargin 4.95

o o o

28 28I _ _ II' II I

[_N Numberof3bichincrementsthatthecontrolbladeiswithdrawnfromfullyinsertedmind

26 ........ 26%

22 , N .... 2o "_, I

16 .... -- 16 ......

1 z |,...... , ,!It

e!!

t , I]4 ' ' jpr

2 ,_,,,; _ = ........ _ _;im _ _--00 0.2 0.4 0.6 0.8 1.0 1_ 1.4 1.6 IJl 2,0 00 2 4 6 | IO 12 14 16 18 20 22 2,1 26 28 30

Core AxialAverageReladvePower Core AxialAverageExposure,GWDIst

1.1.1-14

Figure 1.1.1-15 Equilibrium Cycle Core Data Summary at 6000 MWD/st

i ii liili iL I I I i li i I I il i

II I II II I I I 6000 ICycle Exposure, MWD/st .0I

Cycle E_e:Ny,MWD I032245

....... "- FuU.....my 9Numberof Power s 262.i ill i i inn u

CoreAverageVoidFraction 0.4460IIII [ III I ii III

o ' o ConsFlow,Mib/hr 1.151r MaximumChannelPcaddng 1.3532

0 0 0 Core Axial Power Peak 1.2820.... _L'dOR .....Maximum 0.8823

i iii

-o o MaximumCPI_AT 0.6797nlll,,i ii i

Hot Exc=u R_cfivitT, % N/AI i II I I .........

0 -0 0 Cold Shuldown Margin N/Ali ii

o o'm li

_N Numberof 3 inchIncrementsthatthe controlbladeis wilhdrawnfromfully inserted

2' _ %_,[ 22_o2' __%, II

i_ ,, _

'll ' z II' " tiil _ ,o

, li , tl

' I' ' I4 _ 4 ! ,Oo 0o.4 o.6 o.n t.o 1.2 n.4 1.6 u zo o s 4 6 8 :o _2:4 _6182on24 _ n3o

Core Axial AverageRelative Power Core Axial AverageExposure,GWDlst

I.I.I-15

Figure 1.1.1-16 Falullibrium Cycle Core Data Summary at 8000 MWD/stIIII II I i ii iiiiii [ _ I I IIIII III i[ i mlllllSlll II n IIII I I JI [I I I I

.....,, ....... .....' ,, "76327

Numberof FullPowerDays ,_k_0.6I I II III I

_ Fraction 0.3819core vo.I II I _

0 Cole Row, ldlb/lu" 1.151III III EllHII If[ m|mllm|i

MaximumChmnelPeaking 13895IIIIII _ II I --

_j 12 • 112 C,me AxialPowerPeak 1.2303III I illl ii I _ ]1 I

Maximum_OR 0.8666I nil

0 0 kt_tm_ CI_,_T 0.7_3II IIiiiii i i i iiiiii iiii I

Hot ExcessReactivity,% 0.58Ill i ilUI liililII _ I J

!2 12 Cold ShutdownMargin 4.59ill i i _ I iii iii ii IlUl lull II I

__

0

l L ! I II

B N Numberof 3 inchincremenuthatthecomrolbladeiswithdrawn fullyinsertedfrom

26 ..... 26

24 _- 24

18 , - 18 -

16 16 ,

..... 14

.... n 1_no _,no8 j 8 .....

j,6 6

4 ..... 4

J : ......2 _,d_"" ' ""- "'-

°o o.2 0.4 o.6 o.s _.o _a 1.4 n.6 n.n zo °o s 4 6 o non n4s6nn2o=u 26a 3o

CoreAxialAverageRelativePower CoreAxialAverageExposure,OWD/st

1.1.1-16

Figure 1.1.1.17 Equilibrium Cycle Core Data Summary at End of Cycle

I illIII I IIIIII IIIii ..... IIIIiiiiii II I _111 _ii

I I II ii I I _ III

CycleEzposure,MWDIst- _ I I

Cy_ r_sy, _ 15y_oooI l I II ] II I] IIII I _I_ I

Number of Full Power Days S92 0J I I I . I . _-_ IIIIIIIII I I • 1

cm Aver,eVom_ 0.373sI flit It II + I

Flow,Mlb/lu' 1.151__ .... __ Piing .

I III III1[ III I -- L I IIIII iii II I I I I III I

CornAxial Powa Peak 1.2464

0.'P967iiii I IllII

Maximum_AT 0.7524[ IIII I

_ Ho!a,_ _,cUvi_.,_ 0.0Cold Shuulown Margin 3.90i Illll -. I i i iii I

I II

r I I

I I I I

B N - Number of3 InchIncrementsthatthecomml bladeiswlthdmwn fromfullyinserted

26 r ......

%,.- *_b,.. I24 ...... 24 -

__ _ __. J

,, ..... ,, I. ,, i ,,, . ki

,, !,, |

II

, , + , II, ,_ Ii4

00 ...... 000.2 _4 06 [O i.O il i.l it ii lo 2 4 + i IO nl Ii 16 II lo 22 Ii I ill 1

Core Aria] Average Relative Power Core Axial Average Exvosure, OWI)/st

1.1.1-17

1.1.2 TRANSIENT RESPONSE OF REFERENCE FUEL DESIGN

The StandardSafety Analysis Report(SSAR) for the ABWR was submittedto the NRC in 1987.

The NRC review is approachingcompletion. All majortechnicalissues areresolved (SECY-89-

153 and SECY-90-016) and approvedby the NRC on June26, 1990, Final design approvalwas

expected but was rescheduledfor 1994.

The ABWR was designed for use of generic fuel: Therefore,much of the available behavioral

analyses for the ABWR, includingthe SSAR, are applicableto the presentstudy. In that work,

GE has evaluatedthe entire spectrumof events in nuclearsafety and operationalanalysis areas to

establish the most limiting or design basis events in a meaningful manner. The scope of thesituations analyzed includes anticipated operational occurrences, off-design abnormal

(unexpected) transientsthat induce system operationaldisturbances,postulatedaccidentsof low

probability(e.g., the sudden loss of integrity of a major component), and finally, hypothetical

events of extremely low probability(e.g., an anticipatedtransientwithout the operationof the

entirecontrol rod drive system). In the event analysis, all essential protection sequences were

evaluated until all requiredsafety actions were successfully completed. The event analysis

identifiedfront line safety systems and theiressential auxiliaries,operatoractions, and limits to

satisfy the requiredsafety actions. A partiallist of the events examinedin given below.

• Events that Decrease Core Coolant Temperature

- Loss of Feedwater Heating

- Runout of One Feedwater Pump- Feedwater Controller Failure to Maximum Demand

- Opening of One Bypass Valve

- Opening of al Control and Bypass Valves

- Inadvertent Opening of One Safety-Relief Valve

- Inadvertent RHR Shutdown Cooling

• Events That Increase Reactor Pressure

- Fast Closure of One Turbine ControlValve

- Slow Closure of One Turbine Control Valve

- Pressure Regulator Downscale Failure

1.1,2-1

- GeneratorLoadRejectionwith BypassOn

- GeneratorLoadRejectionwith Failureof One Bypass Valve

- GeneratorLoadRejectionwith Failureof All Bypass Valves

- TurbineTripwith Bypass On

- TurbineTripwith Failureon One Bypass Valve

- TurbineTripwith Failureof All Bypass Valves- InadvertentMSIV Closure

- Loss of CondenserVacuum

- Loss of AC Power

- Loss of All FeedwaterFlow

- FeedwaterPiping Break- Failure of RHR Shutdown

• Events That Decrease Reactor CoolantSystem Flow Rate

- Tripof Three Reactor InternalPumps

- Tripof All Reactor InternalPumps

- Fast Runbackof One ReactorInternalPump

- Fast Runbackof All ReactorInternalPumps

- Seizure of One ReactorInternal Pump

- One Pump Shaft Break

• Reactivity and Power Distribution Anomalies

- Rod Withdrawal Error During Refueling

- Rod Withdrawal Error During Startup- Rod Withdrawal Error at Power

- Control Rod Misoperation

- Abnormal Startup of One Reactor Internal Pump

- Fast Runout of One Reactor Internal Pump

- Misplaced Bundle Accident

- Rod Ejection Accident

- Control Rod Drop Accident

1.1.2-2

• EventsthatIncreaseReactorCoolantInventory

- InadvertentHPCFStartup

The ABWR responseto transients is reportedin Chapter 15 of the SSAR. Response to design

basis accidents is reported in Chapter 6. Finally, compliance with the ASME code for

overpressureevents is reportedin Chapter5. In developing the ABWR two bases were used for

the analysis -- design basis and licensing basis. For the design basis of the NuclearBoiler GE

chose the most conservative core design that wasknown at the time, including the possibility of

the use of mixed-oxide fuel in the future. This was the so-called Core Z design used in the plant_'

design development. In the licensing basis, reportedin the SSAR for limiting transients, GE

used a reference fuel design, typical of today's UO_ fuel offerings, called Core A. Thus, as

alternativemixed-oxide fuel designs used in this presentstudywere developed, GE alreadyhad

• a guide as to which transientsand accidentswould be themost limiting, and whethertheremightbe problemsaccommodatingthe designs.

Table 1.1.2-1 compares the nucleardynamic parameters of Core A, Core Z, and those of the

high burnup mixed-oxide core at the end of cycle, which is the most limiting point for

pressurization transients.

Table 1.1.2-1. Comparison of Nuclear Dynamic Parameters

I ........ hB__Par_eterl Core _, _ i iCore Z Hid ,utnup CoreVoid Coefficient, ¢/% -8.4 -11.6 _ -11.2

....Doppler coefficient, ¢/°C -0.31 -0.43 - -0.63

1.1.2.1 Transient Response

1.1.2.1.1 Determination of the Plant Operating Limit

The purpose of the transient response analysis is to set plant operating limits to avoid the

possibility of departure from nucleate boiling for events expected to occur during the plant

lifetime. For BWRs this has been measured by a parameter called Minimum Critical Power

1.1.2-3

Ratio (MCPR), which is the ratio of the power at which a departure from nucleate boiling is

expected to occur on the hottest rod in a fuel bundle to the bundle's current power level.

Starting from the nominal situation, i.e. MCPR = 1.0, it is necessary to factor in statistical

considerations which account for manufacturing and measurement tolerances, including

uncertainties in the thermal/hydraulic correlations developed by test programs, to measure the

departure from nucleate boiling. This sets what is called the Safety Limit MCPR (SLMCPR).

For the ABWR, this has been calculated to be 1.07, meaning that if the most limiting fuel

bundle in the core is predicted to have anMCPR.of.l.07, there is a 0.1 percent probability that a

rod will have experienced a departure from nucleate boiling. That is the acceptance criterion

approved by the U.S. NRC for BWRs.

Next, all possible transients which can occur (generally due to a single equipment failure or

single operator error) are analyzed to determine the most limiting one in terms of the change ofcritical power ratio which can occur. This so-called ACPR is added to the SLMCPR to

determine the plant Operating Limit MCPR (OLMCPR). The steady-state fuel performance

analysis is then compared to the OLMCPR to determine the amount of operating margin.

In the prior studies of ABWR with Core Z and Core A, GE determined that the most limiting

transient is the Load Rejection with Bypass Failure (LRWOBP) event. Even though this event

involves multiple failures, a long-standing GE-USNRc agreement exists to consider this event

as one of those in Chapter 15 to be compared to the MCPR safety criterion. The primary nuclear

dynamic parameter driving this transient is the void coefficient, since it is the void collapse from

the primary system pressurization which results from suddenly shutting of the steam heat sink

which cases the power rise in the core. In the Chapter 15 analyses, pressurization transients

were analyzed with GE's one-dimensional kinetics code, ODYN, which is capable of following

the effects of the traveling pressure wave through the core as the event progresses. In this code,

basic cross-sections from the three-dimensional steady-state core analysis are used, so it is not

possible to apply conservative multipliers to the key parameters - Void Coefficient and Doppler

Coefficient. Other transients were analyzed using the point model kinetics code REDY. While

this does not compare all transients on an "apples-to-apples" basis, because of the background

studies showing the pressurization transients to be limiting, the OLMCPR will be correctly

computed. The net effect of the Chapter 15 analyses reported in the ABWR SSAR is that

pressurization transients use the dynamic parameters derived from Core A, while the others use

1.1.2.4

the more limiting Core Z. Table 1.1.2-2 shows the results of the three most limiting SSAR

analyses all of which become pressurization-type transients.

Table 1.1.2-2. Results of Transient Analyses

!ii _i i ! _ _ i: i , ! _: _i i i i iD¢i_CdtiealPoWcrR_ti_!i'_,!ii!/!_,,_ili!_ii!i

Generator Load Rejection withFailure of all Bypass Valves 0.10 0.19Fast Closure of One TurbineControl Valve 0.10 0.10Fcedwater Controller Failure -Maximum Demand 0.10 0.16

Comparing the parameters of Table 1.1.2-1 confirms that the LRWOBP event will be the

limiting transient for the limiting mixed-oxide core, based on the background studies.

Therefore, it was the key one analyzed to determine the OLMCPR and available operating

margin for this study. Since two other events also showed comparable Acritical power ratios for

the UO2 core, they were also repeated, using ODYN, for this study. Table 1.1.2-3 lists the input

conditions used for the transient analyses. These are identical to those used in the SSAR.

1.1.2-5

Table 1.1.2.3 Input Parameters and Initial Conditions for S/'stem Response Transients1. ThermalPowerLevel (MWt)

Warranted Value 3926Analysis Value 4005

2. Steam Flow (kg/hr)Warranted Value 7.64x 106

Analysis Value 7.84x 1063. Core Flow (kg/hr)

Rated 52.2x106Maximum 59.0x 106

4. FeedwaterFlow Rate (kg/sec)WarrantedValue 2122

....... Analysis Value 21795. Feedwater Temperature (Celsius) 2176. Vessel Dome Pressure (kg/cm2g) 73.17. Vessel Core Pressure (kg/cm2g) 73.78. Turbine Bypass Capacity (%NBR) 339. Core Coolant Inlet Enthalpy (kcal/kg) 294.110. Turbine Inlet Pressure(kg/cm2a) 69.911. Fuel Lattice N

12. Core Leakage Flow (%) 11.6713. MCPR Safety Limit " 1.0714. Nuclear Characteristics Used in

ODYN Simulations EOEC*

15. Number of Reactor Internal Pumps 1016. Safety/Relief Valve Capacity (%NBR) at

80.5 kg/cm2g 91.3

17. Relief Function Delay (sec) 0.418. Relief Function Opening Time (sec) 0.15

19. Safety Function Delay (sec) 020. Safety FunctionOpeningTime (sec) ....... _. ....._- .....0.3 ...........

1.1.2-6

Table 1.1.2-3 (continued)

21. Set Points for Safety/Relief ValvesSafety Function 82.8, 83.5, 84.2,

84.9, 85.6Relief Function 80.5, 81.2, 81.9,

82.6, 83.3, 88.422. Safety/Relief Valve Reclosure Setpoint-

Both Modes (% of Setpoint)Maximum Safety Limit (used in analysis) 98Minimum Operational Limit 93

23. High Flux Trip (%NBR) 127.524. High Pressure Scram Setpoint (kg/cm:g) 77.725. Vessel Level Trips (m above bottom of

separator skirt)Level 8 (m) 1.73Level 4 (m) 1.08Level 3 (m) 0.57Level 2 (m) -0.75

26. APRM Simulated Thermal Power Trip (%NBR)Analysis Setpoint 117.3Time Constant (sec) 7

27. Reactor Internal Pump Trip Delay (sec) 0.1628. RIP Trip Inertia Time constant for Analysis

(see) 0.6229. Total Steamline volume (m3) 113.2

30. Set Pressure of RIP Trip (kg/cm_g) 79.1

1.1.2-7

Table 1.1.2-4. Sequence of Events for the LRWOBP

_i_i!i_ii_iiii!ii!!i!!iii_i_!i!:ii!!ji!_ii_i_i_iiiiiii_ii_!ii_i_i_i_!iiii!!iEventi,iii_,ii_i_iil!iiii_iii_ii_iii_,_:_: ili,;i:_ii:!ii,__i, :ii,Time (seconds)Turbine-GeneratorDetection of Loss of ElectricalLoad -0.015Turbine-GeneratorLoad Rejection Sensing Devices Trip toInitiate Turbinecontrol Valves Fast closure 0.0

Turbine Bypass Valves Fail to Operate 0.0Fast Control Valve Closure Initiates Reactor Scram and

Tripof Four RIPs 0.0Turbine Control Valves Closed 0.07

Safety/Relief Valves Open Due to High Pressure 1.3Safety/Relief Valves close 7.1Safety/Relief Valves Open Again to Relieve Decay Heat 8.9Safety/Relief Valves Close Again >15.0 (est.)

Table 1.1.2-5. Sequence of Events for the Fast Closure ofOne Turbine Control Valve

Time (seconds)simulate one Main Turbine Control Valve to close 0Failed Turbine Control Valve Starts to Close 0.0

Turbine Bypass Valves Start to Open 2.7Neutron Flux Reaches High Flux Scram Setpoint andInitiates a Reactor Scram 2.84

Water Level Reached Level 3 Setpoint. Four RIP'sare Tripped 7.65

1.1.2-8

Table 1.1.2.6. Sequence of Events for the Runout of all Feedwater Pumps

!ili!:_i!iiiiii'_i!iii!_:_iiiiii_:_!':}I:I,I!i{i__:_,_i__i:ii!_!Ev_n.ti!i,_ !_i:__: z:_i_i i_ ...._ _ _..... , _ Time (Seconds)

Initiate Simulated runout of all Feedwater Pumps (130% at SystemDesign Pressure of 74.9 kg/cm2g on Feedwater Flow) 0Level 8 Vessel Level Setpoint Initiates Trip of Main Turbineand Feedwater Pumps 18.45Reactor Scram and Trip of Four RIP's are Actuated by Stop ValvePosition Switches 18.46

Main Turbine Bypass Valves Opened Due to Turbine Trip ' ' 18.6 .SRV's Open Due to High Pressure . 20.2Safety/Relief Valves Close >25Water Level Dropped to Low Water Level Setpoint (Level 2) >40 (est.)RCIC Flow into Vessel (not simulated) >70 (est.)

Tables 1.1.2-4 to 1.1.2-6 show the key events occurring during the transients and Figures 1.1.2-

1A to 1.1.2-3D graphically display the time variation of key plant parameters. The ACPRs from

these events are given in Table 1.1.2-2. The most limiting event is the LRWOBP as expected;

the limiting delta-CPR is calculated to be 0.19; therefore the OLMCPR would be set at 1.26 for

this core design. Based on the steady-state results reported in Table 2.7-2, the minimum

operating margin would be 23 percent (CPRRAT=0.77 with OLMCPR=I.26 and MCPR=l.63).

GE has typically supplied fuel to currently operating BWRs with operating margins as little as

seven percent; therefore this design is acceptable.

The size of the margin (23 percent) for MCPR is still significantly larger than the 15 percent

margin requirements applied per section 1.1. This is the case even with the handicap of the

additional 0.09 delta-CPR compared to the ABWR SSAR result for the LWROBP transient

(Table 1.1.2-2). This significant margin has been maintained because this plutonium disposition

study is based upon the GE9 fuel design. This fuel design shows significant performance

advantages over the fuel design on which the ABWR SSAR is based, GE8.

1.1.2-9

Ol'+'l'l

PERCENTOF RATED

l.l.2.11 i

_I'Z:'I'I

UNITS

1.0 I _

' ,/// __ ''_'I''__

0.0 I /

1 = VOID REACTIVITY2 = _PPLER REACTIVIi_

M 3 = SC_H REACTIVITY•

M 4 = TOTALREACTIVITY

w

= \

-1.0 \

_ 1- X-2.0 I _ i i ! i I _ i0 1 2 3 4

TIRE (SEC)

Figure 1.1.2-1D Load Rejection With Bypass Failure

I = NEUTRONFLUX2 = PEAKFUELCENTERTENP3 = AVERAGESURFACEHEAT FLUX4 = FEEDWATERFLOW5 = VESSELSTEANFLOM

'150

!

// ..zoo _ •

' ,

! I ! !

0 . . . I . . . .4 8 12 16

TIHE (SEC)

Figure 1.1.2-2A Fast Cl_e of One Turbine Control Valve

1 = LEVEL (INCH REF SEP-SKIRT)Z = g R SENSEDLEVEL (INCHES)3 = N R SENSEDLEVEL (INCHES)

150 4 -- COREINLET FLOM(g)5 = PI.IHPFLOM3 (g)

100

•-. I--,'-', 4z

4

50 ! z

-- _

0 4 8 12 16

TIHE (SEC)

Figure 1.I.2-2B FastClosureofOne Turbine Control Valve

1 = VESSELPRESSURERISE (PSI)2 = STEM LINE PRESSURERISE (PSI)3 = TURBINEPRESSURERISE (PSI)

125 4 = RELIEF VALVEFLOM([)5 = BYPASSVALVE FLOM(Z)6 = TURBINESEAR FLOEI(Z)

6

75 _6

N =

°..

_ 6 • 3 t

.i i" |

-ZS _ _ I I I I I0 4 8 12 16

TIME (SEC)

Figure 1.1.2-2C Fast of One Turbine Control Valve

i

1.1.2-17

1 = NEUTRONFLUX2 ,, PEAKFUELCENTERTEHP3 = AVERAGESURFACEHE;,T FLUX4 = FEEDWATERFLOW5 = VESSELSTEAMFLOg

150

• 4 • 4 4

t .! 'i:1oo_.. i-_.-, I--

,,. b__ C_

F-

ly

!

I

0 I I I I _ I I ! I,0 5 10 15 20

TIHE (SEC!

Figure 1.1.2-3A of All Feedwater Pumps

1 = LEVEL (INCH REF SEP-SKIRT)2 = g R SENSEDLEVEL (INCHES)3 = N R SENSEDLEVEL (INCHES)4 = COREINLET FLOM(%)5 = PUHPFLOH3 (%)150

- \• i..-

# z

3 z

50 .aI

!

I

0

0 5 lO 15 20TIHE (SEC)

Figure 1.1.2-3B Run-up of All Feedwater Pumps

1.1.2-20

.//

./

•" .j"

_ i

1.0o.

-Y• /1 ' o-

, , __t/I't a,tu.u _

.. 1:,o,o,EAcT,v,T, --==_i_,v_ 2 = DOPPLERREACTIVII'Y

.- >- 3 - SCRAHREACTIVITY;., I.-.-

.,_ ,--,> 4 = TOTALREACTIVITYt_" I'"

I,,i,J

-1.0D

$i

I

m

m

I

i

-2.0 , , , , I , , , ,0 5 10 15 20

TIHE (SEC)

Figure 1.1.2-3D Run-up of All Feedwater Pumps

The large MCPR margin for the high burnup mixed-oxide fuel is the result of using the GE9

design.

1.1.2.1.2 Stability

From a safety point of view, the ABWR has provided protection from reactor instabilities which

might cause fuel damage. In BWRs density-wave oscillations have been observed under test and

operating conditions as the reactor power and flow are lowered along the rated rod line. With

this in mind, the ABWR has provided protection by excluding a zone of operation in the power-

flow map (see Figure 1.1.2-4). During power ascension by pulling control rods, if the operator

attempts to increase power above 25 percent with flow less than 40 percent (30 percent pump

speed) an automatic rod block is issued by the control system which prevents further rod

withdrawal until the minimum pump speed is achieved. During transients, if two or more RIPs

are tripped and the flow becomes less than 40 percent with power greater than 25 percent, a

Select Control Rod Run-In (SCRRI) command is issued by the control system to insert rods until

the power becomes less than 25 percent.

The basis of the above protection is to prevent the plant operators from getting the plant into an

operating state where instabilities have been known to occur. In spite of the above protection,

however, the ABWR also monitors the local power through the LPRMs and will issue a scram if

significant oscillations are detected.

The basis of the minimum pump speed limit has been determined from stability analyses

demonstrating a 95 percent confidence that oscillations will not occur at or above this point.

From many tests and analyses, GE has determined the primary drivers for instability. Once

again, the void coefficient is the most significant parameter for the purpose of the mixed-oxide

studies. Therefore, stability limits were checked to see if further restrictions in the operating

map would be necessary.

Stability is calculated by the use of frequency-domain codes which linearize the basic dynamic

equations for small perturbations and report the behavior of the reactor in terms of a second-

order feedback system. Commonly, a decay ratio is reported; this is the ratio of the second over

shoot to the first overshoot of the reactor response to a step change in input. Thus a decay ratio

1.1.2-22

<1.0 means the reactor is stable, and a decay ratio >1.0 means the reactor is unstable. In

calibration of GE's methods to test data, a 20 percent conservatism is needed to achieve a 95

percent confidence of avoiding instability; thus it is required to predict the decay ratio <0.8. The

least stable point in the operating domain of Figure 1.1.2-4 is the intersection of the rated rod

line with the minimum pump speed line. Thus, stability of the high burnup corewas calculated,

using the same procedures and methods documented in the SSAR for Core Z. The most limiting

void coefficient for the mixed-oxide core was taken from the mid-cycle point (-12.9

cents/percent, per Figure 1.1.1-11).

The results are shown in Table 1.1.2-7. It can be seen that at the current minimum speed (30

percent), i.e. core flow of 40 percent the decay ratio = 0.86. Therefore, a second case was run

with a slightly higher speed in order to determine how much further restriction would be

necessary. Referring to Table 1.1.2-7, it can be seen that a nominal increase in required

minimum pump speed will satisfy the stability criteria. This will not present any significant

operating restriction.

Table 1.1.2.7 Stability Evaluations_ii_Core Decay.Ratio l _:! i il Hot Channel,Deca_,Ratio

30% Speed/40%Flow 0.86 I 0.3935% Speed/45%Flow , 0.73 I 0.17Design Criteria <0.80 I <0.60

1.1.2.1.3 Anticipated Transients Without Scram (ATWS)

ATWS events are a special class of transients also analyzed in Chapter 15 of the SSAR. Because

of the low probability of occurrence, ATWS acceptance criteria are somewhat different thanthose for normal transients. These include:

a. Core coolability, determined by demonstratingfuel temperatures <2200Fahrenheit

b. Nuclear Boiler peak pressure not to exceed service Level C of the ASME Boiler

Code, i.e. 1500 psig.

c. Containment pressure not to exceed s_rvice Level A, i.e. 45 psig

1.1.2-23

i i i

I I I I I i I i I I i130,,-

TEN OF TEN INTERNAL PUMPS OPERATING -"

PERCENT PUMPSPEEO120 .-- ll_m3m_ ,x, .-.

0 0 NATURAL CIRCULATIONI 30

110 -- 2 40 ._3 5O4 60 7 8

100 S 70 100% POWER - 3926 MWt tt 1¢_ 6 _

6 80 ,,jr_7 90 10095FLOW - 115.1 mlb/hr

9O 8 99 4 -100% SPEED = 1500 rpm

PERCENT ROD LINE 3.lU

oo "I x ,O I_GI ON !11 2;.. a. B 100i,_ t-- FOR MOX F1EL-. z 70 C 8O B _

m D 60Ua: E 40 REGION IVm 60 F 20is. C

0 REGION III50

I

40 l J!REGION I I

3OE

#I REGION II

20 l

TYPICAL F STEAM SEPARATOR LIMIT10 STARTUP /

PATH /

0 v0 10 20 30 40 50 60 70 80 90 1O0 110 120

PERCENT CORE FLOW

Figure 1.1.2-4 Modified Power-Flow Operating Map for ABWR with MOX Fuel

In the SSAR, ATWS analyses were reported showing significant margins to the acceptance

limits. However, background studies for ABWR have shown that the peak pressure predicted

for ATWS pressurization transients is within 20 psid of that for the ASME overpressure

transient (reported below). Thus ATWS is not a limiting overpressure basis for ABWR.

1.1.2.2 Overpressure Analysis

In BWRs the number of safety/relief valves is normally set by requirements of the ASME code

for overpressure protection. This analysis is a special transient in which the MSIVs are suddenly

closed, the normal MSlV position scram is assumed to fail - the reactor scrams on high neutron

flux. In addition no credit is given for safety/relief valves opening in the power-actuated relief

mode; credit is given only for the spring mode. The acceptance criterion for this analysis is

Nuclear Boiler peak pressure less than service Level B of the ASME code, i.e. less than 1375

psig. Since this is an overpressure transient, the void coefficient will once again play a

significant role in the results. Therefore, this analysis was repeated for the limiting mixed-oxidecore.

Results show that the peak vessel pressure is <1272 psig, well within the acceptance criterion.

Even for a further delayed case using the second backup scram (high pressure scram) the result

is 1284 psig.

1.1.2.3 Accident Analyses

Response to design basis accidents is reported in Chapter 6 of the SSAR for Loss-of-Coolant

Accidents (LOCAs) and in Chapter 15 for Reactivity accidents. Reactivity accidents, such as

Rod Drop or Rod Ejection, which are dominated by the fuel Doppler Coefficient, are eliminated

in the ABWR by design features which preclude their happening. For the ABWR no analyses

were reported in the SSAR; therefore no further consideration was given to this class of events

for the present study.

1.1.2-25

1.1.3 FUEL CHARACTERISTICS AFTER IRRADIATION

The decay heat rate on a per bundle basis is shown in Figure 1.1.3-1. The average isotopic

content of the fuel bundles for the reference spiking case is shown in Table 1.1.3-1. The

isotopics are calculated five days after reactor shutdown.

Table 1.1.3.1. Average Bundle Isotopic Content (grams per bundle)Plutonium S )ent Fuel Case

Th-232 tr Kr-83 7.463 Eu- 153 24.808U-233 tr Ru-1Ol 168.020 Eu-154 9.074U-234 5.820 Cd-ll3 2.494 Eu-155 3.465U-235 382.964 Pm-148G 0.085 Gd-154 18.709U-236 144.963 Pm-151 0.002 Gd-155 0.094U-238 168000 Ru-103 6.601 Gd-156 332.846,,,,. .,,,.,,

Pu-239 1414.876 Sm-148 24.027 Gd-157 0.103Pu-240 1185.369 Xe-133 0.686 Gd-158 390.969, ,,.,

Pu-241 492.230 1-135 tr Tb-159 14.568Pu-242 223.414 Pr-143 1.936 Gd-160 210.204Pa-231 tr Rh-103 116.481 Nd-145 115.947

Np-237 27.778 Rh-105 0.023 Gd tails 3.704Pu-238 31.542 Cs-133 222.921 Fiss.Prod.

Balance 2378.898Np-239 2.321 Xe-135 trAm-241 28.525 Nd-143 133.578

,, ,,

Am-242 tr Tc-99 139.838Am-243 43.571 Xe-131 90.899,,. ,,,

Th-228 tr Pm- 147 29.849Th-230 tr Pm-148M 0.243Cm-242 10.085 Pm- 149 0.035Cm-244 13.062 Sm-147 17.438

Sm-149 0.519.,,

Sm-150 54.718, ,,,,

Sm-151 3.011Sm-152 34.573

tr - less than 0.0005 grams/bundle

1.1.3-1

Figure 1.1.3-1 Decay Heat Load/'or the Pu Spent Fuel Case

00000 ....

10000_ .... ,

iooo

wattsper __-_ ___._ __._..___bundle

00 ,, ii l

I0 500 lOGo 1500 2000 _ 3000 3500 4000 4500

• daysaftershutdown

1.1.3-2

1.2 ALTERNA_ CORE DESIGNS FOR PU DISPOSITION

The reference fuel design for discharging Pu as spent fuel will be the design reportedin the

previous section (Section 1.1) of this Phase 1C report. This design maximizes Pu utilization

with a core averageenrichmentof 3.48% and a dischargeexposure of 37 GWD/MT (typical of

the GE9 commercial assemblydesign). As reportedin Phase 1A, thisdesign requiresno change

to the reactor system. The fuel material has been well proven with Uraniafuel and requires

minimal developmenteffort. Fordispositionin 25 years (fromproject inception), 6 reactorsare

required,with the first reactorcoming on-line at the end of 7 years and each subsequent reactor

coming on-line every six months. The study assumes a 75% capacity factor. If the capacity

factoris the ABWR reference87%,only 5 reactorswill be required.

Due to the large capital cost required for the reference scenario, other options are described

which employ fewer reactors. It is important to note that the disposition time is the single most

influential variable in lowering the required number of reactors for disposition. Three cores are

described in the following sections. The first option is the reference spent fuel core, described

in Section 1.1. This option gives the most complete utilization of the excess Pu. The second

option utilizes a core design incorporating a 5% enrichment assembly and an exposure of 37

GWD/MT (with a possible redesign to accomodate 30 GWD/Mt). The third option is a core

design which employs 3.75% Pu enrichment and is capable of up to 30 GWD/MT (detailed for

Phase IA in Section 2.6).

The major parameters for the Option 2 core design which employs a 5% Pu enriched assembly

and a burnup of 37 GWD/Mt are given below in Table 1.2-1.

1.2-1

Table 1.2.1. Equilibrium Cycle Key Parameter Summary,

........... High EnrichmentCycleLength,EFPD ...........................392Cycle Energy, GWd 1,539Cycle Exposure, MWd/metrictons .' ..', 98.6.!Core Mass,metric tons 149.1i i i i ,

ReloadEnrichment,W/0U-235 5.0ReloadBatchSize,bundles..... 232,, ,,,,,,ii iIll,I I ,,, ,

Maximum MAPRAT 0.72-Maximum CPRRAT 0.75

Maximum LHGR, Kw/ft(LHGR limit= 14.4) I0.2MCPR ( OLMCPR = 1.25 ) .............. 1.67Minimum Cold ShutdownMargin 3.9HotExcess Reactivity at BOC 4.0

1.2.1 Alternatives for 100 Metric Tons in 25 Years

There are two major routesto disposing the 100MT of Pu in 25 years. One is to utilize the same

fuel design as the reference design and lower the exposure, another is to increase the enrichment.

Several different core design studies have been conducted which show both approaches are

possible. The following options are shown graphically in Figure 1.2-1.

Option 1 uses the Reference Spent Fuel core which is 3.5% enrichment and 37 GWD/MT

burnup. This design maximizes Pu utilization. Six reactors are required, with the first reactor

coming on-line at the end of 7 years and each subsequent reactor coming on-line every six

months. The study assumes a 75% capacity factor. If the capacity factor is the ABWR reference

87%, only 5 reactors will be required.

There are two possibilities for Option 2 (5% enrichment). Option 2A has 5% enrichment and a

burnup of 37 GWD/Mt. This design would require only 4 reactors for disposing 100 MT in 25

years at 75% capacity factor. Option 2B would use a redesigned 5% assembly to give a burnup

1.2-2

of 30 GWD/Mt and faster Pu disposition capability. This design would require only 3 reactors

for disposing 100 MT in 25 yeats at 75% capacity factor.

Option 3 utilizes the core design which employs 3.75% Pu enrichment and is capable of up to

30,000 MWD/MT. This design would require only 4 reactors for disposing 100 MT in 25 years

at 75% capacity factor.

1.2-3

100 Mt of Pu in 25 Years

Figure 1.2-1 Options for 100_t Pu Disposed in 25 Years

1.2.2 Alternatives for 50.100 Metric Tons in 25 Years

The system design andcore designs will be the same as the reference spent fuel and based on the

options outlined under the previous section. For disposing less than 100 MT in 25 years, fewer

reactors will be needed. These alternates are discussed below and shown in Figure 1.2-2.

Using Option 1 ( reference spent fuel) and 75% capacity factor, 50 MT of Pu could be disposed

using 3 reactors, 66 MT using 4 reactors and 83 MT using 5 reactors, in 25 years as shown in

Fig. 4. In all these cases Pu is utilized to the fullest.

The Option 2A design (5% enrichment and 37 GWD/Mt burnup) could dispose 50 Mt with 2

reactors or 75 Mt with 3 reactors. With the Option 2B fuel design (5% enrichment and 30

GWD/Mt burnup), 2 reactors could be used to dispose 50 MT of Pu in 21 years or 66 MT could

be disposed in 25 years.

Option 3 (3.75% Pu enrichment) could dispose 50 MT with 2 reactors and 75 MT could be

disposed using 3 reactors. Pu is not utilized as fully as with the Reference Fuel Design. Note

that Option 3 gives the same results as Option 2A. This will be discussed more fully in Section1.3.

1.2-5

50-100 Mt of Pu in 25 Years

mjl

I o " i

. Dwi i

I " |$ ..............................._................................................................................•................ _..............r'""_4" ,..................

i j, " : i

i , ,,_" •..io" Ii• _ _ !

0 i m "a"

w

0 ! ! [ _ _" i i [ i .

=1 [_... _ i ;__-,-'-- _ i i i [ i | i

_1_-_-----_--_ _................ L.............................. ' ......... '.---................ -- ................... '--..........................

" _ _ _ .... Option 1

i _ "" Options 2A & 3....................................... _ ............. _............................. .._ . = ..................

1 i - i ..... [.............. i .............................- ................... _ ............................. Option 2Bi : i i i i

: i

O

50 55 60 65 70 75 80 85 90 95 100

Mass of Pu Disposed (Mt)

Figure 1.2-2 Options for 50-100 Mt Pu Disposed in 25 Years

1.2.3 Alternatives for 100 Metric Tons in More Than 25 Years

The system designand core designs will be the same as the reference spent fuel and based on the

options already discussed for other alternatives. For disposal times of greater than 25 years,

fewer reactors will be needed. These alternates are discussed below and shown in Figure 1.2-3and 1.2-4.

Using the Reference Spent Fuel assembly with 3.48% Pu enrichment (Option 1), 2 reactors will

dispose 100 MT in 48 years of operation or 1 reactor could be used for disposing 50 MT during

this sa_aeperiod.

Option 2A (5% assembly enrichment, 37 GWD/Mt) will dispose 100 Mt in 36 years with 2

reactors or 50 Mt during the same 36 years with 1 reactor. Option 2B (5% assembly enrichment,

30 GWD/Mt) will dispose 100 MT of Pu in 54 years with one reactor. The design lifetime of

the ABWR is considered to be 60 years. Alternatively, 50 MT could be disposed of in 27 yearswith one reactor.

Using Option 3 fuel design (3.75% assembly enrichment), 100 MT could be disposed of in 36

years with 2 reactors or 50 MT during the same period with 1 reactor.

1.2-7

Figure 1.2-3 Options for 100 Mt Pu Disposed in Greater Than 25 Years

100 Mt of Pu, Variable Years

1. , OpUonI (

z _--[ _-- °PU°"2Bl/_!/ / /O ,f/ , V /

36 48 54

Number of Years

Figure 1.2-4 Options for 50 Mt Pu Disposed in Greater Than 25 Years

50 Mt of Pu, Variable Years

,./k/. / / / /./

Option 2A &'$ :iOption 2B . Optionl

I Option 3• r

i

,.... / / .//0-27 36 48

Number of Years

1.2-8

1.3 RELATIONSHIP BETWEEN Pu ENRICHMENT, DISCHARGE EXPOSURE,

DISPOSITION TIME, ISOTOPICS AND NUMBER OF REACTORS

This section is a generic discussion of the relationships that govern the meaningful factors in Pu

disposition. These relationships represent all reactors in the fission option.

1.3.1 RELATIONSHIP BETWEEN QUANTITY TO BE DISPOSED, TIME,

ENRICHMENT AND EXPOSURE:

There exists a direct relationship between the quantity of Pu that needs to be disposed,

disposition time, Pu enrichment, exposure and the number of reactors necessary for the

disposition process. By defining any three of these variables, the fourth will be predefined.

Since the reactor system represents a major initial (investment) cost, it is useful to explicitly state

this relationship although it is based on rather elementary principles.

Variables:

N: Number of reactor years available for disposition (does not include

time necessary for building and testing reactors)

Q: Quantity of Pu to be disposed (Mt)

X: Average assembly Pu enrichment (%)

E: Average discharge exposure (MWd/Mt)

C: Capacity factor

P: Reactor power level (MWt)

# of Reactors = Q*E ,

(0.01)*X% * N years *365 days/year * C * P

For the ABWR and the current Pu disposition guidelines:

C = 0.75 P = 3926

1.3-1

# of Reactors = O Mt * E Mwd/Mt ,

10747*X%* N years

The above equationis shown in the following:

MWd producedby reactor in given time = N years * 365 days/year* C * P MWt

Total heavy metal mass (Mt) = O Mt .0.01 *X

Total MWd from disposing given mass = 0 Mt * E MWd/Mt.0.01 *X

- The relationship between the various parameters is graphically shown in Figures 1.3-1 through

3. The number of reactors required is lowered with lower discharge exposure or with higher

enrichment or longer disposition times. It should also be noted that with lower number of

reactors, the amount of electricity produced (and therefore the revenue stream) is also lower but

the capital cost is also diminished..

The curves show hypothetical cases. The feasible design space for each variable is a subset of

the total system. As an example, a core design with 3.48% Pu enrichment and 37,000

MWD/MT exposure which satisfied all the requirements was reported in Phase 1A. Such a fuel

design would require 6 reactors for disposing 100 MT of Pu within 25 years from project

inception (or 18 years of operation). If the value of Pu enrichment were increased to 20% and if

the discharge exposure could be kept at the same value, the above relationship would indicate

that only one reactor is required. However, such a design with very high Pu enrichments

appears unfeasible in an LWR without extensive system level changes and an extensive re-

examination of the licensing basis. Even if such high Pu enrichment designs were feasible, they

may require extensive and unpredictable development costs but more importantly, as discussed

below, the spent fuel isotopics are none the better for their "high exposure" compared to a lower

Pu enrichment design where the fuel is discharged at a lower exposure.

1.3-2

100 Mt of Pu at 40 GWD/Mt

• J _! 1 J

6 , ; ...................;..... i ,i J

"J i i i iS "-....................t..... .... i I I _

% _ . . !

¢_ No l_,dditional _" ...._ ..........: ............-_...............................]-.........................i............................_..i................. _ ,.-- erable_ s . .,.! Kequires -'11<. _ Requires Consa ' : I

'_ _ Develo )ment Requ:red[ ' t • ,'% , Moderate i Developmen_: : i !_.............. ......._ .......................... Development i & Licensing Re, ,iew3 N ..............*- _,-....-i ........................T.........................I..............................................

E= , \! . ,

2 '_ i' _ ..................4J.......................--.'.-i-'..-,,.-,..-..:..._-.-,i.........................I _ ! "-.. !

I ! _ ! _ " " ,, ,, ..._,,,, ,...................................

i : i i _-

I i I _ i0

0 2 4 6 8 l0 12 14 16 18 20

Pu Enrichment (%)

Figure 1.3-1 100 Mt of Pu at 40 GWD/Mt

100 Mt of Pu at 30 GWD/Mt

Figure 1.3-2 100 Mt of Pu at 30 GWD/Mt

100 Mt of gu at 20 GWD/Mt

Figure 1.3-3 100 Mt of Pu at 20 GWD/Mt

1.3.2 RELATIONSHIP BETWEEN SPENT FUEL ISOTOPICS, Pu ENRICHMENT,AND FUEL EXPOSURE

It is generallyrecognized thatPu separatedfrom any LWRspent fuel could be ,_sedfor making

weapons, even though it may differ greatly from so-called "bomb-grade"material. A clearer

definition of the resistanceof the spent fuel to subsequentuse for weapons is needed and thiswould include the degree of difficulty in working with irradiatedfuel, the extent to which

diversion is possible without detection, the minimum number of assemblies that need to be

divertedfor a critical mass of plutonium, and the deterioration/shelf-lifeconsiderationsfor Pu

derived from spent fuel. The following discusses the isotopics of the spent fuel undervarious

core design options.

Pu is destroyed according to energy produced. For any one reactor design, a given number of

reactors will destroy a given amount of Pu over a given time period. Pu disposed out of

stockpile can change with different core enrichments and different burnups. Consider two

different core design options, one that uses a high Pu enrichment of 7% and high exposure of

40,000 MWD/MT and another with a lower enrichment of 3.5% and 20,000 MWD/MT. These

two core designs, based on the discussions of the earlier section, would require the exact same

number of reactors to dispose of a given quantity of Pu in a given disposition time. Although

the first design, in view of its "higher" exposure appears to better "utilize" the fuel and hence

lead to better denaturing, in fact the fuel isotopics at discharge is just about the same. This is

shown in Fig. 1.3-4. As a first approximation, assume the number of reactors and disposition

time are fixed and assume the range of fuel enrichments that could be used in an LWR without

considerably affecting the licensing basis (1 to 10%). For this case, the extent of Pu denaturing,

that is (Pu239 + Pu241)/Total Pu, could be expected to be very nearly the same for various core

designs. Isotopic compositions of the spent fuel do vary, but the extent of "denaturing" is aboutthe same.

If the number of Pu disposition reactors are restricted, using a higher.enrichment and higher

discharge exposure has a secondary advantageous effect. The MOX throughput needed for

higher exposure fuel (expressed in MT of MOX fuel) is lower compared to the needs of a lower

enrichment fuel which is discharged at a lower exposure. One item to remember is that beyond

1.3-6

about 5% Pu enrichment levels, additionaldevelopments/confirmatory tests may be required asthe fuel material deviates more and more from the conventional urania fuel.

1.3-7

Figure 1.3-4 Plutonium Isotopic Comparison

Pu FissileFractionI._

Oe_

Pu

0.4IIIIII i

0.3 [-------- MOx(3.5_)0.2 i-'- "" Mo,_.o_jI

O.1 BWR SpentFuel ExposureRange

00 5 10 15 20 25 80 S5 40 45 50

Exposure (MWD/ST_

1.3-8

2.0 FUEL CYCLE

2.1 MIXED OXIDE FUEL FABRICATION REQUIREMENTS FOR SPENT FUELSCENARIOS

The fabrication requirements for the fuel pins and bundles for each of the ABWR spentfuel scenarios being investigated are summarized in Table 2.1-I. (Note: detailed fuel pindesigns are given in Section l.O on Core and System Performance). Table 2.1-I identifiesand summarizes the amounts of the key strategic materials needed in each type of fuelbundle. The strategic materials are uranium, plutonium and gadolinium. The total amountsof each strategic material in each bundle is given. These materials are distributed amongthe 60 pins in each bundle.

Table 2.i-1 also summarizes the nominal heavy metal throughput requirements for the fuelfabrication plant for the ABWR spent fuel scenarios. In each scenario the fuel fabricationcampaign is planned to begin three years before the first reactor begins operation. Thedesign operating cycle length is 523 days (392 days at full power and 131 days formaintenance and refueling outage). For these multi-reactor scenarios, after the first reactor,another reactor is planned to begin initial operation every half cycle (262 days) until allreactors have started up. The duration of the campaign needed to completely transformeither 100 MT or 50 MT of weapons-grade plutonium into ABWR fuel bundles is alsoindicated in Table 2.1-1. The mixed oxide fuel throughput for the fabrication plant design,as indicated in Table 2.1-1, is specified to provide approximately 15 percent excessproduction capability.

2.1-1

,,, ,,,

Table 2.1-1

Characteristics of Mixed Oxide Requirements for the Various ABWR Spent FuelScenarios

, " ,,1, fm "" ' 'n,T i' ,,, ! i "' f ii ,, fl,'lJ , iI , ,i ,,,,, , , ,,,,, L ', , , ," ,' ,',", i

SPENT FUEL SCENARIOS

Characteristics .... 1 I 2 3 4

Plutonium, MT 100 100 50 50

Reactors, # 6 2 3 1

Time to Dispose of Pu, years 25 60* 25 60*

ABWR Fuel Bundle Design:Fuel Pins, # 60 60 60 60

Uranium, kg 172.7 172.7 172.7 172.7Plutonium, kg 5.3 5.3 5.3 5.3Gadolinium, kg 1.0 1.0 1.0 1.0

Fabrication Requirements (Material Usage):Uranium, MT/yr 168 56 84 28Plutonium, MT/yr 5.2 1.7 2.6 0.8Gadolinium, MT/hr 1 0.3 0.5 0.2

AsssemblyRequirements:MOX Fuel Pins, #/yr 58,320 19,440 29,160 9,720Bundles, #/year 972 324 486 162Campaign Duration, yrs 19.2 58.8 19.2 58.8

Fabrication Plant Design'MOX Fabrication, MT/yr 220 80 110 40

*ABWR design lifetime

2.2 MOX FUEL HANDLING AND DISPOSAL

2.2.1 Criticality Analyses for Storage, Handling & Repository

This section summarizes the criticality safety analyses conducted for theplutonium-basedfuel bundles. Both the freshfuel and spent fuel bundleswere includedinthe criticalityanalysisof shippingcontainer. In addition,preliminaryinvestigationof spentfuel storagein the repositorywas conducted.

2.2.1.1 Criticality Analyses for Fuel Shipping

For more than 20 years, the RA series shipping containershave been used byGeneralElectric Nuclear Energy (GENE) to ship BWR fuel elements to domestic andinternational customers. Extensive analyses have been performed to demonstratecriticalitysafetyof these containersfor the transportof a wide rangeof 7 x 7, 8 x 8, 9 x 9,and 10 x 10 BWR fuel assemblies. The analysis performed in this report focusesspecificallyon the GE9 bundledesignwith MOX fuel, as used in the referencespent fuelalternativefor Pu disposition.

2.2.1.1.1 Criticality Safety Requirements

There are two principal criticality safety requirementsfor shipping container: (1)classification as a Fissile Class I shipping container, as documented in 10CFR71 -Packaging and Transportation of Radioactive Material (Reference 2.2.1-1), and (2)qualification under the 1985 IAEA Regulations for the Safe Transport of RadioactiveMaterials (Reference2.2.1-2). These requirementsare brieflysummarizedin the materialwhich follows.

10CFR71 Requirements for Fissile Class I _;hinnint Container

The criticality safety requirements for a Fissile Class I shipping container, asdocumented in 10CFR71,are as follows:

"A Fissile Class I package mustbe so designed and constructed and its contents solimitedthat:

(a) Any number of undamaged packages would be subcritical in anyarrangement and with optimum interspersed hydrogenous moderationunless there is a greater amount of interspersed moderation in thepackaging, in which case the greater amount may be assumed for thisdetermination,and

(b) Two hundred fiRy (250) packages, if each package were subjected to...Hypothetical Accident Conditions ... would be subcritical if stackedtogether in any arrangement,closely reflected on all sides of the stack bywater, and with optimuminterspersedhydrogenousmoderation."

In addition, it is required that a single Fissile Class I container be subcritical withoptimumhydrogenousmoderationand when closely reflectedon all sidesbywater.

The HypotheticalAccidentConditionsreferredto hereare:

2.2.1-1

1. a 30 foot (9.15 m) freedroptest

2. a I meter free dropand puncturetest

3. a thermal exposure or fire test in which the container is exposed to800 °C for at least 30 minutes,and

4. an immersion test equivalent to at ! "_ cC _,-_ (15.25 m)of water for at least 8 hours.

IAEA Requirements for ArraYsof Shipping PackaEes

In 10CFR71, it is noted that unlimitedarraysof undamagedcontainers and 250unit water reflected arrays of damagedFissile Class I containers were requiredco bedemonstratedtc be subcritical ("Damaged" refersto the worst case condition resultingfrom the "Hypothetical Accident Conditions".) The designationof shipping containersinto "Fissile Classes" is differentfrom the current IAEA regulations which specify genericrequirementsfor arraysof containersas follows:

"An array of packages shall be subcritical. A number"N" shall be derived: =....assumingthat if packageswere stackedtogether in any arrangementwith the stack :....

closely reflectedon all sides by water 20 cm thick (or its equivalent)both of thefollowing conditionswould be satisfied:

a) Five times "N" undamaged packages without an_ing between thepackages would be subcritical;and

b) Two times "" _qagedpackages with hydrogenous moderationbetweenpackages tc nt which results in the greatest neutron multiplicationwould be suL _."

If unlimitedarraysof both undamagedand damagedpackages are demonstratedto............... --be subcritical, individual shipments of such packages are not required to be limited in

numberdue to criticality safety, but will be assigned a TransportIndex due to exposureratesmeasuredat the container.

2.2.1.1.2 Description of RA Series Shipping Containers

For more than 20 years, the RA series shipping containers have been used byGeneral Electric Nuclear Energy (GENE) to ship BWR fuel elementsto domestic andinternational customers. The RA series containers consist of rectangular steel innercontainers transportedin wooden outer overpacks. The wooden overpack containers aredesignedwith ethafoam and honeycombcushioningbetween the metal inner container andthe inside walls of the outer. The inner metal conta/ner has two internal ethafoam-cushioned channel sections each of which can hold a single fuel assembly.

The origin_ designed RA-1 inner container was modified in the 1970's toaccommodatea longer fuel assembly. This was accomplished by adding a largerend capto the body of the inner. The new designwas designated as the RA-2. Subsequently, andas a result of consideration for fabrication and handling, the longer bodied RA-3 (with a

2.2.1-2

shorterend cap) was introduced. Correspondingchanges in the outer wooden containerwere also madeto accommodatethe new innerdesigns.

The RA-3D inner metal container is constructed of stainless steel 321 with aminimum16-gaugeouter shell andstructuralandreinforcingcomponents. Insidethe innercontainer,•there is an innerbasket formed of two perforatedmetal channels. The inner..

basket is held inplace by the six 3 inch by.3 inch by 1/8 inch (7.62 cm by 7.62 cm by0.3175 cm) thick angled supports welded to the innerwall of the outer shell. Within theinnerbasket, fuel assembliesrest on additionalethafoamcushioning. At the upperend of

' the innercontainera removableend cap is attachedby.boltswhich screw,into threaded_.=_bolt holes welded onto the main body. The innercontaineris sealed by a lid and rubbergasket which are held in place by 14 stainless steel clamps. A pressurerelief valve isinstalled on the inner containerwhich is designed to pass up to 2 cfm (56.6 l/m) of air ifthe pressuredifferentialbetween the inside and outside of the containerexceeds 0.5 psi(3450 Pa).

The outer containeris a rectangularwooden box 33 inches high by 32 inches wideby 207 inches (83.82 cm by 81.28 cm by 525.78 cm) long. It is fabricatedof 2 inch by 4inch (5.08 cm by 10.16 cm) wooden studs,wood planks,and plywood and is lined with8.5to 9.0 inch (21.59 cm to 22.86 cm) thick phenolicresin impregnatedhoneycomband 3 ....to 4 inch (7.62 to 10.16 cm) thick ethafoampads. Cutoutsare madein the ethafoam andhoneycomb to accommodatethe handles and liftinglugson the innercontainer._SubjecttomeetinS the minimumpackagerequirements,the ethafoamand honeycombcushioningareotherwise arranged in the outer container to minimize vibrationaleffects on the fuelassembliesbeingtransportedin the innercontainer.

During the packagingand handlingof the RA-3D container,one or more BWRfuel assemblies are placed in the chambers in the inner container. (If only one fuelassembly is packed, the other chamber is usuallyfilled with a dummybundleto providebalanced loading.) Prior to being placed in the inner containerfor shipment,each fuelassembly is first preparedby installingplastic separatorsbetween rows and columnsof thefuel rods and byenclosingthe entirefuel assemblyin a thin plastic sheath.

The RA series shippingcontainersare currently licensed as Fissile Class I shippingpackages for the transport of 7 x 7, 8 x 8, 9 x 9, and specific 10 x 10 BWR fuelassemblies. The RA-3D shippingcontaineris also currently licensedas a Type-Apackagein accordancewith the Regulationsfor the Safe Transportof Radioa_ive Materials,1985edition (Supplement1990) of the InternationalAtomicEnergyAgency (IAEA) for generic9 x 9 BWR fuel assemblies.

2.2.1.1.:$ Analytical Technique

In this analysis,neutronmultiplicationfactors(k_'s or kefl_s)have been calculated

with the GEMER.4 (Reference 2.2.1-3) and the MCNP (Reference 2.2.1-4) Monte Carlocode. The following sections providebrief descriptionsof these two codes.

2.2.1-3

The GEMER.4 Code

GEMER.4 is an enhanced combinationof the geometry modeling capabilitiesofthe well knownKENOMonte Carlocode and the sophisticatedcross section handlingandneutrontrackingof GENE's IV_RITMonte Carlo code. MERIT is a derivative of theBattelle Northwest BMC code, and is characterizedby its explicit treatment of resolvedresonance in material cross section sets. The MERIT treatment uses cross sectionsprocessedfrom the ENDF/B-IV library.These cross sections areprepared in a 190 broad

.... groupformatand the groups inthe resonanceenergyrangehave the formof Breit-Wigner-resonance parameters. These parameters are used in explicit sampling to determinethe

value of the cross section at the neutron's energy. Since resonances are consideredexplicitly,fluxweighting of cross secamnsis unnecessaryand only one cross section set isrequired per isotope (and per temperature). Thermal scattering of hydrogen in water,paraffin, etc., is represented by the S(oql3)kernels in the ENDF/B library. The types ofreactionsconsidered in the Monte Carlo calculationsare fission, elastic scattering, inelasticscattering, and (n, 2n) collisions. Absorptionis implicitlytreatedby reducing the neutronweight through determiningthe non-absorption probabilityat each collision.

The geometry treatment in GEMER.4 includes the regular and generalizedgeometry options from the KENO-IV code and anenhanced complex embedded option ......_...which permitsgrouping of regular geometry regions inside of other such regions. In the.regular geometry treatment, a geometric configuration is generated by defining boxeswhich when stacked together in one, two, or three dimensional arraysmake up the totalmodel. Within each box, individualregions and their corresponding materials (limited toone material per region) are defined using nested simple geometry forms such asCYLINDERs, SPHEREs,and CUBOIDs. Within each box, each region must completelyenclose all previously defined regions except that successive regions may share commonboundaries. The GEMER.4 geometry package permits arraysof boxes to themselves beenclosed by the simple geometry forms so that modeling a close water reflectorcan beachieved byusing a simplewater filled CUBOID to surroundthe array.

In the generalized geometry treatment, geometric modeling is achieved using theequations for quadratic surfaces and by specifying the various materials that lie in theregions bounded by the surfaces. This option allows a description of very complicatedgeometry models, but it becomes cumbersomeand computationally inefficient when largenumbersof surfaces (such as would be required for a lattice of fuel rods) are necessary.Generalized geometry boxes can, however, be stacked in (three dimensional) arrayswithregular boxes.

In the complex embedded geometry treatment,arrays of regulargeometry boxesmay be placed inside of one or moreother boxes. For example, ifBox Type 1 describes afuel rod, Box Type 2 a Gad rod, Box Type 3 a water rod, and Box Type 4 the regionbounding an entire fuel assembly,then Box Types 1, 2, and 3 may be embedded in BoxType 4 to give the complete descriptionof the assembly. Box Type 4 may then itselfbeassembled into an arraywhich can then be surrounded by regions representingpackagingor water reflection.

2.2.1-4

The three geometry options describedabove may be used in any combinationtogeneratea geometrymodel. In the presentanalysis, the regulargeometry optionhas beenused to describethe outer regions of the shippingcontainer,and the complex embeddedoption has been used to describe the fuel assemblies and their immediatelyadjacentregions. A reflectiveboundaryconditionwas used to analyzeinfinitearraysof undamaged

' . and damaged,containers,while tile individual (inner container)geometry models werestackedusing the box arrayoption.

• The GEMER.4 code has beenvalidated by_comparison,against more than one._.hundredcritical. experiments. •.These critical experimentshave included a ..signiticant._..numberinvolving comparable lattices of fightwater reactortype low enrichedfuel rods.

"Fromthisvalidation,GEMER.4'sbias has been conservativelyestimated to be -0.003 forthe range of materials,water-to-fuelratios, and kefl_sand koo'sapplicableto this analysis.The minus sign in this value indicates that the neutron multiplicationfactors areunderpredicted.

This bias is a result of three primaryfactors. The first is the randomstatisticaluncertaintiesin the Monte Carlo calculationsof the benchmarks themselves. Typicalrandom errors (a) for these calculations are in the range of 0.0Ol to 0.005 andconsequentlythe averageof N such values has anuncertaintyon the order of 0.0005 to0.002 (o divided bythe squareroot of N).

The second significantcomponent of the bias is in the uncertaintyof the crosssection sets. As noted above, GEMER.4uses a single unique cross section set for eachisotope and hence the benchmarkcalculationsalso serve to benchmarkthe cross sectionsets. This uncertaintyin the cross section data is probablythe largestcontributorto thebias.

The thirdidentifiablecontributorto the bias is the cumulativeeffect of other code

and modelinglimitations.Theseincludeuncertaintiesdue to programmingapproximations,the broadenergygroup cross section structure(as opposed to the cross section data setitself), and inherentlimitationsof the Monte Carlomethoditself. The contributionto thebias of these types of errorsis the smallestof the three types, especiallysince these errorsare normallysmall and manyof them will average out as the diversity and numberofbenchmarkcritical experimentsincreases.

The MCNP Code

MCNP is a general-purpose, continuous-energy, generalized-geometry, time-...dependent,coupledneutron/photonMonte Carlotransportcode. It solves neutralparticletransport problems and may be used in any of three modes: neutrontransport only,photon transport only, or combined neutron/photontransport, where the photons areproducedby neutroninteractions. The neutronenergyregime is from 10"11MeV to 20MeV, and the photon energy regime is from 1 keV to 100 MeV. The capabilitytocalculatek-effective eigenvaluesfor fissilesystemsis also a standardfeature.

MCNP uses continuous-energynuclear data libraries. The primary sources ofnucleardata are evaluationsfrom the EvaluatedNuclear Data File (ENDF) system, the

2.2.1-5

EvaluatedNuclear Data Library(ENDL) and the ActivationLibrary(ACTL) compilationsfrom LivermoreNational Laboratory,and evaluations from the Applied Nuclear Science(T-2) Group at Los Alamos National Laboratory. Evaluated data are processed into aformat appropriate for MCNP by codes such as NJOY. The processed nuclear datalibrariesretainas much detailfromthe original evaluationsas is feasible.

.,

.... • _ Nuclear-data tables exist for neutron interaction, photon-interaction, neutrondosimetryor activation, and thermalparticlescatteringS(ot,13)kernels.._Over.500 neutron• .

_" " '.'interactiontables are available for approximately 100 --differentisotopes-ror,elements............. .-:-.Photoninteractiontables exist for all.elements fi'om.Zfl-through Z---94.•,Thedata in the ,.

-_ _ photon interactiontables allow MCNP to account for coherent and incoherent scattering,...... _:photoelectric absorptionwith the possibilityof fluorescent emission, and pair production. ,

Cross sections for nearly 2000 dosimetryor activationreactions involving over 400 targetnuclei in ground and excited states are part of the MCNP data package. These crosssections may be used as energy-dependent response functions in MCNP to determinereaction rates. Thermal data tables are appropriate for use with the S(c_,13)scatteringtreatmentin MCNP. The data include chemicalbindingand crystallineeffects that becomeimportant as the neutron'senergy becomes sufficientlylow. Data are available for lightand heavy water; berylliummetal, berylliumoxide,.benzene, graphite,,polyethylene,,and .,_zirconiumand hydrogenin zirconiumhydride.

2.2.1.1.4 Modeling

Modelint of the Inner Container

A geometrymodel of the RA-3D innercontainer is shown in Figure 2.2.1-1. Themetal in the shell, the innerbasket, and the angled basket supports are all includedand aremodeled as stainless steel. The perforatedinnerbasket is includedby modelingit as metalwith a reduced density (85% of the normal stainless steel density). The ethafoamcushioning between the fuel assembly and the innerbasket as well as the plastic sheathingaroundthe fuel assemblies are not included. Eliminatingthis internalmoderatingmaterialis conservativefor the following reasons:

1. Arrays of undamaged containers are over-moderatedby the cushioning materialand wood in the outer container. Therefore,the omission of moderatingmaterialswill result in increasingthe calculatedkoo'S.

2. For accident condition arrays, the-fire test (which is part.of the .Hypothetical_.Accident Conditions) completely burns away all internal flammable materials(except plastic separators), Even were this not the case, the accident arrays areanalyzed with interspersedmoderationwithinthe innercontainerwhich is variedtodetermine the optimum amount. The presence of additional ethafoam or theplastic sheaths around the assemblieswill therefore not result in greater koo'sbut

will only cause a slight change in the optimuminterspersedmoderatordensity.

2.2.1-6

Dimensions in inches

Fuel Rod Cladding Basket is 85% DensityZirconium Stainless Steel

Fuel Within Fuel Rod Inner Container and Supports areCladding Full Density Stainless Steel

Polyethylene Separators Other Open Areas areBetween Fuel Rods interspersed Water

Figure 2.2.1-1 RA-3D Inner Container Geometry Model

ModelinEof the Outer Container

The outer containeris modeled as shown in Figure 2.2.1-2. This model is alsoconservativesince it does not includeall of the moderatingcushioningmaterialsthat areactuallypresentin the package. Note in particularllhatportionsof the regions betweenthe innerand outer containersare empty (i.e. void) in the model. The model corresponds

.... in this regard to what is known as the "Minimum Packaging Model':.,.The Minimum.....PackagingModel also includesa50% reduced materialdensity of the ethafoamto permit.......some flexibilityin the arrangementof the cushioning.

Modelin_ of the Fuel Assembly

In the actual bundle design for the referencePu spent fuel alternative,there areseven distinctfuel rodtypes each with differentPu enrichmentandGad loading. Namely,the 28 peripheralrods,all without Gad,include4 difl%rentPu enrichments(4 with 1.0%,8with 1.6%, 8 with 2.3%, and8 with 2.8%);while the 32 interiorrods all contain 1%Gadandconsist of 4 rodswith 2.8% Pu, 8 with 3.2%, and 20 with 4.2%. Thus, the averagePu enrichmentis 2.057% and 3.775%, respectively, for the 28 peripheralrods and 32interior rods. The bundleaveragePu enrichmentis 2.973%. All rods contain 0.71% ofU-235.

A general practice in criticality safety analyses is to carry out k-effectivecalculations with a simplifiedfuel assemblymodel that contains only one or two fuelenrichments in the bundle. The purposes of this approach are (a) to simplify thecalculation and, more importantly,(b) to provide criticality safety results that would beapplicable to a spectrum of fuel bundle design that falls within the enrichmentspecifications. Otherwise, time-consumingcriticality safety analysiswould have to beperformedwhen a new fuel elementdesignis conceived.

The fuel assemblymodel used in this study is simplifiedto two enrichmentzones.The 28 peripheralrods were assumedto contain no Gad and have a Pu enrichmentof2.8% while the 32 interiorrods have a Pu enrichmentof 4.2% and a Gad enrichmentof1%. All rods contain0.71% of U-235.

Fuel assemblieshave been modeled in complex embeddedgeometry. The modelconsists of constructinga box correspondingto each type of unit in the fuel assemblyandembeddingthese units in anotherbox with the samedimensionsas the fuel assembly. Thefollowing assumptionsand parametershave beenused in this scheme:

1. The diameterof the fuel regionwithin the fuel rodshas been"assumedto equalthenominalinsidediameterof the Zirc cladding(0.419 inches). For determinationofthe fuel nuclide densities, the fuel pellets have been assumedto have a density ofthe maximum theoretical value of 10.96 g/cm3, which is greater than typicaldensityfactorsfor realUO2 pellets. The resultingdensity is then averagedover theinside of the claddingby applyinga smeardensity of 0.9622 (square of ratio ofpellet outerdiameter,0.411 inches,overcladdinginnerdiameter).

2.2.1-8

Dimensions in inches

m Ethafoam i I Void

_._ Honeycomb Wood

Figure 2.2.1-2 RA-3D Outer ContainerGeometryModel

2. Thethicknessof the clad was assumedto be the nominal(0.032 inches).

3. An active fuel length of 150 inches (381 cm) was assumedfor the MOX and Gadrodsfor all cases.

4. The plasticinsertswereconservativelymodeledascylindricalshells(0.020 inchthick) surroundingeachindividualfuel rod suchthatthe amountof plasticwasgreaterthan thatactuallypresent. The plasticshells extended over the full length.....of the fuel rods.

5. The gadoliniumdensity was calculated as the specifiedrpercentage of the MOX =_density. To add conservatism, the specified weight percent of gadolinium isfurthurreducedby 7.5% in the model. For example, for 2 weight percentgad,1.85 percentwas actuallyused. The displacementof MOX by gadolinium wasneglected. The averaged(or "smeared") fuel nuclidedensitiesused in this analysisare listedin Table2.2.1-1.

6. All structuralcomponents in the fuel assembly except for the cladding wereconservativelyignored.

Table 2.2.1-1Nuclide Densities (Atoms/Barn-Cm) For Fresh MOX Fuel Rods

Nuclide 2.8%Pu_o Gad 4.2% Pu/l% Gad

U-235 1.6572E-04 1.6572E-04U-238 2.2237E-02 2.1914E-02Pu-239 6.0386E-04 9.0580E-04Pu-240 3.6617E-05 5.4926E-05Pu-241 1.9272E-06 2.8908E-06O-16 4.6090E,02 4.6556E-02Gd - 3.1275E-04

2.2.1.1.5 Analytical Procedure

The criticality safety criteriafor shippingcontainers meeting the requirements.forFissile ClassI containers and the 1985 IAEA Regulations can be summarized as follows:

1. !ndiv.idual undamaged container: An individual undamaged container must besubcritical when optimallymoderated and fullyreflected by water.

2.2.1-10

2. Infinite array of undamaged containers: An infinite array of undamagedcontainers must be subcriticalwith optimum interspersed moderation betweencontainers.

3. Array of damage4,containers:

3a. FissileClassI: An arrayof 250 containerseach subjectto the HypotheticalAccident Conditionsmust be subcritical when.closely reflected by water ..andwhen arrangedin the most reactiveconfiguration.

3b. 1985 IAEARegulations: An array.of" 2N" containers,each subjectto theHypothetical Accident Conditions, must be subcritical,when closelyreflected by water and when arrangedin the most reactive configuration.The numberN sets the TransportIndex and is defined as the allowablenumberof containerswhichmay be transportedin any shipment.

It can be seen that, for the arrayof damagedcontainers,if the allowablenumberofcontai:_,_rsto be shipped is taken to be infinite,then it is clear that the requirementsfor Fissile Class I packages is a subset of the IAEA Regulations set ofrequirements. Therefore,requirement3b is used for the analysisof infinite arrayofdamaged containers.

Based on these requirements,compliance with the FissileClass I requirements andthe 1985 IAEA Regulationsare typicallydemonstratedin the following manner:

1. A single container is demonstratedto be subcritical by showing that two fuelassembliesare subcriticalwhen optimallymoderatedand fullyreflectedbywater.

2. An infinite arrayof accident condition containers is demonstratedto be subcriticalin the following steps:

a. The accident condition container is defined based on prior HypotheticalAccident Condition tests of real containers as consisting only of the innermetal container with all cushioning and burnable components removed.The absence of the sealing gasket between the lid and body of the innercontainermeans that water in-leakage mustbe considered.

b. To show that infinite arraysof damaged containers with various densitiesof interspersedwater insidethe container is subcritical.

3. An infinite arrayof undamaged containersis demonstratedto be Subcritical in the ..following steps:

a. To show that the undamaged container is over moderated bydemonstratingthat (1) placing water between the undamaged contair_ersdecreasesthe kooof the arraysand (2) placing water within the containers

decreasesthe kooof the arrays.

b. To show that an infiniteclose packed arrayof the undamaged containersissubcritical.

2.2.1-11

Extensive criticalitysafety analyses performedpreviouslyfor BWR fuel bundles inthe RA series containers have shown that the accident arrayshave significantly highermultiplicationfactors than any of the undamaged container arraysor the single containerand hence are the limiting cases in the criticality safety analysis. Therefore, only theaccidentarraysare investigatedin this study.

2.2.1.1.6 Results for Fresh Fuel Bundles

This section documentsthe resultsof criticalitysafety analysis.for.an infinite_array.of damaged containers. Fresh fuel bundles shown in Section 2.2.1.1.4 were used in the -calculation of neutronmultiplication.

Forthe RA series containers,the HypotheticalAccident Conditionsresulted in thewooden outer container and all of the internal ethafoam, honeycomb, rubber,and plasticbeing burnedaway. With the rubbersealing gasket in the innercontainergone, in-leakageof water during the immersiontest was a certainty. With the destruction of all of theburnable materials, arraysof damaged containers are no longer over-moderatedand theaddition of interspersed water may cause the array koo to increase. Other than thedestruction of the burnablematerials, the HypotheticalAccident Condition tests did notresult in any changes in the fuel assemblies or the inner container that could havesignificant impact to the criticalitysafety results. Minor changes in geometry due to thedrop test and the fire test actuallymade it moredifficult ratherthan easier to achieve theclose-packed accident arraysthat are assumed in criticality safety analyses. There was, ofcourse, no loss of neutron absorbingmaterials in the inner metal container or in the Gadrods. Froma criticality safety perspective, the key issues are water in-leakage (which isassumed to be optimum anyway) and damage to the fuel assemblies or inner containerleading to a morereactive configuration.

The results of the analysis are presented in Tables 2.2.1-2 and 2.2.1-3 based onMCNP calculations. These results are for the accident condition container arraysfor therod configurationwhich contains 1%Gad in the interior rods. Table 2.2.1-2 shows arrayneutron multiplicationvalues as a function of varying interspersedmoderatordensity forthe case with Pu enrichments of 2.8% and 4.2% in the peripheral and interior rods,respectively. The results indicate that the damaged arrayis significantlysubcriticalfor allmoderator densities. The maximum kQois found to be 0.7881 nominal, or 0.7903

includinga 2o uncertainty,at a moderatordensity of 0.075.

Table 2.2.1-3 shows array neutron multiplication values for the case with auniform Pu enrichment of 4.2% in all fuel rods. As expected, the increase of Puenrichmentfrom 2.8% to 4.2% for the peripheralrodsraises the neutron multiplicationbyabout0.04 - 0.06, dependingon the moderatordensity. However,the results indicatethatthe damaged array is also significantly subcritical for all moderator densities. Themaximumkoois found to be 0.8406 nominal, or 0.8430 includinga 20 uncertainty, at a

moderatordensity of 0.075.

2.2,1-12

Table 2.2.1-2Neutron Multiplication Factors from MCN-P

for Infinite Arrays of RA-$D Containers in Accident Condition ........(Fresh Bundle, 2.8/4.2% Pu Enrichment, 1% Gad)

FractionalWater Nominal 1oDensity K-infinity Uncertaintyo.000 0.6668 0.00110.050 0.7834 0.00110.075 0.7881 0.00110.100 0.7816 0.00120.125 0.7671 0.00120.150 0.7525 0.00131.000 0.5431 0.0041

Table 2.2.1-3Neutron Multiplication Factors from MCNP

for Infinite Arrays of RA-3D Containers in Accident Condition(Fresh Bundle, 4.2/4.2% Pu Enrichment, 1% Gad)

FractionalWater Nominal 1oDensity K-infinit.x Uncertainty0.000 • 0.7064 0.00110.050 0.8289 0.00110.075 0.8406 0.00120.100 0.8405 0.00110.125 0.8328 0.00120.150 0.8247 0.00121.000 0.6062 0.0023

Comparison Between MCNP and GEMER.4

Tables 2.2.1-4 and 2.2.1-5 show a comparisonof neutron multiplicationvalues asa function of varying interspersed moderator density between the MCNP results andGEMER.4 results. For the case with 2.8%/4.2% Pu enrichment,MCNP shows lowerneutron multiplication factors than GEMER.4. The differences range from 0.01 at amoderatordensity of 0.150 to 0.04 at a moderator density of 0.050. The case with auniform4.2%Pu enrichmentalso shows similarpattern. The cause for these differencesis

2.2.1-13

not clear at this time. More detailed study is needed to understandand resolve thedifferences.

Table 2.2.1-4Comparison of Nominal Neutron Multiplication Factors

Between MCNP and GEMER.4for Infinite Arrays of RA-3D Containers in Accident Condition

(Fresh Bundle, 2.8/4.2% Pu Enrichment, 1% Gad)

FractionalWaterDensity MCNP GEMER.40.050 0.7834 0.81790.075 0.7881 0.8138O.1O0 O.7816 O.79830.125 0.7671 0.78210.150 0.7525 0.7635

Table 2.2.1-SComparison of Nominal Neutron Multiplication Factors

Between MCN]Pand GEMER.4for Infinite Arrays of RA-3D Containers in Accident Condition

(Fresh Bundle, 4.2/4.2% Pu Enrichment, 1% Gad)

FractionalWaterDensiW MCNP GEMER.40.050 0.8289 0.86830.075 0.8406 0.8633O.100 0.8405 0.85400.125 0.8328 0.83750.150 0.8247 0.8208

2.2.1.1.7 Results for Spent Fuel Bundles

This section documentsthe resultsof criticalitysafety analysis for an infinite arrayof damaged containers for spent fuel bundles. These calculations were performedusing ageometrymodel which is similarto the one used for the analysisof freshfuelbundles. Theprincipal differences are (1) the removalof Gad (burnout in spent fuel), (2) change of Pucompositions that reflect the designed fuel bumup of 38,000 MWd/t, (3) inclusion ofbundle channel (0.1 in thick) and, (4) no plastic separatorbetween fuel rods.

The results of the analysis are presented in Table 2.2.1-6 based on MCNPcalculations. These results are for the accident condition container arraysand show array

2.2.1-14

neutron multiplicationvalues as a function of varying interspersedmoderatordensity forthe case with the spent fuel bundle that has as-builtPu enrichmentsof 2.8% and 4.2% inthe peripheraland interior rods, respectively. The results indicatethat the damaged arrayis significantlysubcriticalfor all moderator densities. The maximum_ is found to be

0.7652 nominal,or 0.7696 includinga 20 uncertainty, at a moderatordensityof 0.100.

Table 2.2.1-6Neutron Multiplication Factors from MCNP

for Infinite Arrays of RA-3D Containers in Accident Condition(Spent Fuel Bundle, 2.8/4.2% Pu Enrichment at Fresh, No Gad)

FractionalWater Nominal 1oDensity K-infinity Uncertainty0.000 0.5487 0.00360.050 0.7325 0.00280.075 0.7588 0.00250.100 0.7652 0.00220.125 0.7598 0.00270.150 0.7491 0.00331.000 0.6145 0.0029

2.2.1.1.8 Conclusion for Fuel Shipping

The criticalitysafety requirementsfor classification as a Fissile Class I shippingcontainer and for qualificationunderthe 1985 IAEA Regulations for the Safe Transport ofRadioactiveMaterials have been appliedto the ILA-3Dcontainerwith Pu-based GE9 8 x 8fuel assemblies. These have included requirements for the subcriticality cf singleundamaged containers, infinite arrays of undamaged containers, and infinite arrays ofdamagedcontainers.

Foran infinite arrayof damaged containers,it has been shown that when optimallymoderatedbywater, the system of such packages is subcriticalwith the specifiedPu-basedGE9 fuel assembly for Pu disposition application. The maximum koo including 2ouncertaintyis 0.79 for fresh fuel bundlesand 0.77 for spent fuel bundles.

Previous criticality safety analyses have shown that accident arrays havesignificantlyhigher multiplicationfactors than any of the undamaged container arraysorthe single containerand hence for the RA-3D are the limitingcases in the criticalitysafetyanalysis. Therefore,although analyses for a single RA-30 container and an infinite arrayof undamaged containers were not performedin this studs, meeting the criticality safetyrequirements for these conditions is not perceived to be a problem based on priorexperience.

L

2.2.1-15

2.2.1.2 Criticality Analyses for Spent Fuel Storage in Repository

Monte Carlo calculationswere performedon discretelymodeledspent fuel storagearrays to determinethe eigenvalue of the Yucca Mountain repository filled with spent

I MOX fuel containers in normalandaccident(water flooded)conditions.l

A spent fuel storage container is designedto hold 10 BWR fuel assembliesfor longterm repository storage. The container is a 15 foot steel drum with a 200 mil wallthickness and an outside diameterof 28 inches. It is buried verticallyin the ground, withadjacent containers spaced 15 feet away in a respective tunnel. Individual tunnels arespaced 126 feet apart. Figure2.2.1-3 shows a cross sectional view of the proposed single.,.containergeometrywith 10 BWR fuel assembliesthat was used in this analysis.

The spent fuel assemblies are discretely modeled for the criticality analysis,incorporatingend-of-life heavy metal isotopics. The plutonium content of the averagespent bundleis depleted 35%, relativeto the average fresh bundle,with appreciablyhigherburnout of Pu-239, and appreciablebuildupof Pu-240.

The numeric models used reflectiveboundaryconditions on the sides and bottomof the container arraysto simulateinfinitetunnels, both in length and number. The waterflooded scenario incorporates a 10 foot tall waterregion above the containers.

The 10 bundle container array generates approximately 3500 watts of heat 10years at_erfinal irradiation, when the bundles are installed in the repository for long termstorage.

The normal environment (non-flooded) Monte Carlo analysis yields a repositoryeigenvalue of 0.288 ± 0.003 (1-sigma). The accident (water flooded) environmentanalysisyields a repository eigenvalu_of 0.896 ± 0.002 (1-sigma). Replacing the water inthe central water rod position of each of the ten assemblies in each container with B4Cyields a water floodedrepository eigenvalueof 0.790 ± 0.003 (1-sigma).

The comprehensive Monte Carlo criticality analysis indicates that long termrepository storage of spent fuel MOX assemblies, utilizing the indicated repositorygeometry scheme, poses no criticalityconcerns.

2.2.1-16

References for Section 2.2.1

2.2.1-1. Title 10, Code of FederalRegulations, Part71, United States of America.

2.2.1-2. "Safety Series No. 6, Regulations for the Safe Transport of RadioactiveMaterials, 1985 Revised Edition_, published by the International AtomicEnergy Agency (IAEA) Vienna, Austria.

2.2.1-3. • "GEMER.4 User'sManual,',J. T. Taylor,GE Nuclear Energy, November .1989.

2.2.1-4. "MCNP u-A General Monte Carlo Code for Neutron and PhotonTransport," J. F. Briesmeister,Editor, LA-7396-M, Los Alamos NationalLaboratory,September 1986.

2.2.1-18

2.2.2 ACCEPTABILITY OF SPENT FUEL FROM Pu DISPOSITION INREPOSITORY

I. Introduction

The Nuclear Waste Policy Act of 1982 made DOE responsible for developing an

underground repository for the highly radioactive waste from civilian and DOE sites.

Amendments to this act in 1987 directed DOE to investigate only the Yucca Mountain site

for this repository. Currently the utilities have signed the standard contract given in 10CFR

Ch.III Part 961, Standard Contract for Disposal of Spent Nuclear Fuel (Reference 1).....

Article VI.A.1 of this contract provides the criteria for spent fuel acceptance. The

applicability of this contract as stated in Part 961.2 extends to "spent nuclear fuel or high-

level radioactive waste, of domestic origin, generated in a civilian nuclear power reactor."

It is assumed tha: the spent fuel generated by the disposition program will qualify under the

category of "spent fuel generated in a civilian nuclear power plant" and the rest of this

section briefly examines whether the general criteria for acceptance of spent fuel are

satisfied when MOX fuel is used instead of the standard urania fuel.

The general criteria for acceptance given in Article VI of 10CFR Part 961 (which in turn

refers to Appendix E of this Part), as well as the information which the purchaser of the

contract is to provide, are specified in this and succeeding Articles. These criteria have

been evaluated for potential applicability to spent fuel deploying MOX fuel. It has been

found that all the general criteria are met by the spent fuel to be discharged from the

disposition program.

II. General Criteria for Disposal:

The general criteria stated in Appendix E of 10CFR Part 961 is attached as Appendix B _f

this (Phase 1C) report. The spent-fuel from the disposition program will meet all the

requirements enunciated for "standard fuel" of this Appendix. These criteria do not place

any limit on the fuel bumup. Appendix B of this report also contains data on the discharge

exposures of spent fuel, both those already discharged as well as potential fuel discharges.

The average discharge exposure at this time is close to 30000 MWD/MT while the expected

average for future discharges in the year 2010 is 36000 MWD/MT.

All waste forms to be dibposed of must be extensively characterized so that their behavior

during disposal is understood. Currently, there is an extensive characterization program

ongoing that will continue for several years on LWR urania fuel. It will be necessary to

2.2.2-1

show that the MOX fuel behaves the same way. Ongoing studies include oxidation,

dissolution, cladding and gaseous release. A preliminary assessment for these areas is

given below.

III. Waste Acceptance System Requirements

DOE document DOE/RW-0351P (Reference 2) issued by the Office of Radioactive Waste

Management in January 1993 and Yucca Mountain Site Characterization Project Change

Directive CR No. DCP-060, dated 2/5/93 (Reference 3) describe the functions to be

performed and the technical requirements for a Waste Acceptance System for accepting

spent nuclear fuel (SNF) and high-level radioactive waste (HLW) into the Civilian

Radioactive Waste Management System (CRWMS). As a starting point, it is worthwhile

noting that the only significant difference between the standard spent fuel from urania

fueled assemblies and those from the MOX fueled disposition reactor, consists of the

higher fraction of transuranics, specifically Pu isotopes, in the discharged bundle. The

decay heat for up to 300 years is dominated by the abundance of fission products which in

turn depend upon the bundle exposure. Thus, the decay heat is essentially the same for the

urania fueled bundle as for the MOX fueled bundle taken to the same exposure.

Table F1.1.1 of Reference 3, reproduced in Appendix B, defines the Waste Acceptance

Criteria. These requirements have been reviewed in detail and only the following criteria

are considered to be affected by the presence of higher amounts of transuranics:

10 CFR 60.135 "(a) High-Level Waste Package in general.

"(1) Packages for HLW shall be designed so that the in situ chemical, physical, and

nuclear properties of the waste package and its interactions with the emplacement

environment do not compromise the function of the waste package or the performance of

the underground facility or the geologic setting.

(2) The design shall include but not be limited to considerations of the following

factors: solubility, oxidation/reduction reactions, corrosion, hydriding, gas generation,

thermal effects, mechanical strength, mechanical stress, radiolysis, radiation damage,

radionuclide retardation, leaching, fire and explosion hazards, thermal loads, and

synergistic interactions."

10 CFR 60.43 "License conditions shall include items in the following categories:

(1) Restrictions as to the physical and chemical form and radioisotopic content of the

radioactive waste"

2.2.2-2

10CFR60.131 "Criticality control. All systems for processing, transporting,

handling, storage, retrieval, emplacement and isolation of radioactive waste shall be

designed to ensure that a nuclear criticality accident is not possible unless at least two

unlikely, independent, and concurrent sequential changes have occurred in the conditions

essential to nuclear criticality safety. Each systems shall be designed for criticality safety

under normal and accident conditions. The calculated effective multiplication factor (Keff)

must be sufficiently below unity to show at least a 5% margin, after allowance for the bias

in the method of calculation and the uncertainty in the experiments used to validate themethod of calculation.

10CFR 72.124 "Criteria for nuclear criticality safety. (a) Design for criticality

safety. Spent fuel handling, packaging, transfer, and storage systems must be designed to

be maintained to be subcritical and to ensure that, before a nuclear criticality accident is

possible, at least two unlikely, independent, and concurrent or sequential changes have

occurred in the conditions essential to criticality safety. The design of handling, packaging,

transfer, and storage systems must include margins of safety for the nuclear criticality

parameters that are commensurate with the uncertainties in the data and methods used in

calculations and demonstrate safety for the handling, packaging, transfer and storage

conditions and in the nature of the immediate environment under accident conditions.

Of the requirements listed above, 10CFR 60.131 and 10CFR 72.124 have been

addressed in a previous section of this report where it is shown that subcriticality is

maintained for proposed geometries under consideration for repository disposal. 10 CFR

60.43 and 10 CFR 60.135 requirements primarily pertain to making sure that the

engineered barrier system (EBS) performs as intended. Containment of the waste within

the EBS is expected to be complete for 300 to 1000 years and limited to less than 1 part in

105 per year of the 1000 year inventory beyond this period. The nuclides of interest in this

regard are 14C, 85Kr and 3H and short lived isotopes such as 90Sr, 137Cs during the

containment period and 99Tc, 129I, and 135Cs during the post-containment period. Of

these, Sr, Cs, Tc and I are of interest because they are expected to move to the fuel grain

boundaries or the pellet-cladding gap region. The inventory of these isotopes should be

comparable for MOX fueled and conventional urania fueled discharges for the same level of

exposure although there are small differences in the fission product yields from Pu.

Typical urania fueled assemblies could contain up to 1% Pu at discharge. While the Pu in

the spent fuel from MOX fueled assemblies would be higher, there are considerable data

from MOX fuel developed for Liquid Metal Fast Breeder Reactors which indicate that the

movement of these fission products is not correlated with Pu enrichment. Therefore, the

2.2.-.-3

release rates of spent fuel from MOX fueled assemblies should be no worse than that fromthe conventional assemblies.

IV Other Disposal Criteria

Heat load from a single waste container (or canister) would appear to be a key

criterion for disposal. With regard to spent fuel, the repository is likely to be required to

package and dispose of a variety of fuel configurations. Container materials would have to

be selected based on material properties as well as its corrosion characteristics. The higher ....

Pu content of the PDR-SNF is judged unlikely to affect the corrosion characteristic of the

container material. A key parameter in this regard is the decay heat load of the containerand the manner in which the waste form will be distributed within the container. One

approach, proposed by LLNL, would seek to keep the emplacement hole walls above the

unconfined boiling point of water in the unsaturated zone (about 97C at the repository

elevation) while others have proposed keeping the package as cool as practical. In any

case, it is likely that the spent fuel will be "packaged" within the outer container to achieve

the required objectives. While "geometric tailoring" of the waste products (SNF) within

the container to achieve these objectives has been proposed, "receipt tailoring" or managing

the waste inventory to obtain the required levels of waste heat has also been suggested. In

the latter case, a single container may contain not only PDR spent fuel SNF waste but other

SNF as well. By such combination of measures, it will be possible to maintain the heat

load within adequate limits.

Studies and field work are still ongoing and the set of requirements for the

acceptability a given spent fuel form or composition might undergo changes in the future.

The design information for the repository given below was taken from DOE document

DOE/RW-0198, Nuclear Waste Policy Act, Site Characterization Plan Overview, Yucca

Mountain Site, USDOE-OCRWM, December 1988:

Repository Total Area 2100 Acres

Effective Repository Area 1400 Acres

Repository Heat Limit 57 kW/Acre

Container Separation along Tunnel 15 feet

Separation between Tunnels 126 feet

Container Area allocation 1890 sq. ft.

Number of Containers per acre 23

Heat Limit per container (average) 2.5 kW

2.2.2-4

However, a recent private communication (T. Doring, B&W, (702-794-1857) indicated

that the heat load requirement is undergoing further review and has not been fixed at thistime.

The PDR-SNF assemblies will be able meet the design constraints above equally as well as

conventional urania fueled SNF assemblies. No specific requirement on the allowable

quantities of actinides or fission products per container was found.

V. Spent Fuel Characteristics from Pu Dispositign

The isotopics of the spent fuel from spent fuel for the reference case where the fuel is taken

to an exposure level of 37,000 MWD/MT are discussed in Section 1.2.3 of this report.

After interim storage for 10 years prior to its placement in the repository, the heat load is

mainly due to long-lived fission products. The heat load per assembly as a function of time

following discharge is also discussed in Section 1.2.3. Ten years after discharge, the heat

load per assembly is less than 400 watts. Based on this figure and assuming that storage

commences immediately after 10 years of storage, a maximum of 6 assemblies could be

stored in each container. This is not different from the disposal scheme proposed for

typical commercial spent fuel as the heat load contribution is principally from the fission

products which are related to the exposure. The added transuranic actinides, principally in

the form of the additional Pu in the assembly (compared to typical commercial spent fuel)

does not add materially to the heat load.

VI. Conclusions

Based on the information available to-date, it is judged that the spent nuclear fuel

assemblies from the disposition reactor could be.stored in the permanent repository and that

all the applicable requirements will be met. The slightly higher transuranics in the spent

fuel does not preclude its disposal in the permanent repository.

References

1. 10CFR Ch.III Part 961, Standard Contract for Disposal of Spent Nuclear Fuel

2. DOE document DOE/RW-0351 P, issued by the Office of Radioactive Waste

Management, January 1993.

2.2.2-5

3. Yucca Mountain Site Characterization Project Change Directive CR No. DCP-060,

dated 2/5/93

2.2.2-6

D 2.2.3 SPENT FUEL PROLIFERATION RESISTANCE

I. Introduction

One of the work scope elements for this phase of the study called for an assessment of the

ability of the reference spent nuclear fuel (SNF) to meet proliferation resistance

requirements. A set of criteria/requirements in this regard is yet to be developed. Work

was however initiated in anticipation of a set of criteria becomingavailable and this section

summarizes the results of these evaluations.

II. Proliferation Resistance through Various Steps in the Disposition

Process

It was pointed out in Phase 1A studies that the proposed disposition option

•increases proliferation resistance incrementally during each stage of the disposition process:

a. Destruction of the pit shape - Requires reworking the materialb. Conversion to Oxide - Requires stripping the oxygenc. Downblending with Urania powder - Requires separation of the Uraniad. Sintered to pellet form - Urania separation is more difficulte. Loaded into fuel pins and assemblies - More difficult to divert in view of the sizef. Dispositioned as SNF - Requires extensive infrastructure, remote handling

Table 1, taken from DOE Order 5633.3A shows that attractiveness level of special

nuclear materials in various physical and chemical forms and it is seen that the

attractiveness level decreases progressively through the disposition process.

III. Safeguardability of SNF

We judge the safeguardability and diversion resistance of the spent fuel from the

disposition, process to be comparable to the SNF from conventional urania fuel for the

following reasons:

(a) Both conventional SNF and SNF from the PDR will require an extensiveinfrastructure, remote handling and processing equipment to separate theplutonium.

(b) Diversion of dispositioned SNF or commercial SNF are both easily detectable.

(c) As illustrated in Figure 1, for exposures higher than about 10000 MWD/MT, thePu isotopics of the SNF from disposition is virtually the same as SNF fromconventional urania fuel when the Pu enrichment of the MOX fuel is about 3.5%.Above this enrichment level, to achieve the same isotopic "degradation" wouldrequire proportionately higher levels of fuel exposure. These results from Figure 1are not expected to differ significantly between different LWRs.

2.2.3-1

Pu Fissile Fraction

239pu+241p UJL_

Pu

_iI uo_ " '!_ MOx(3.5%)......M°'_7"°'_l

O"I BWR Spent Fuel Exposure Range, i , I , I,, I ,, I I , ,,i _ I, •

0 5 10 15 20 25 30 35 40 45 50

Exposure (_D/S'O

Figure J,, Plutonium Isotopic Comparison

It should however be pointed out that no criterion governing the dependence of

proliferation resistance on Pu isotopics is available. It is also not clear that a basis for such

a criterion exists in view of the significant fission cross-sections of all Pu isotopes for high

energy neutrons.

The only aspect where the SNF from conventional urania fuel and MOX fuel SNF

differ slightly concerns the amount of Pu in a single assembly. Typically, high burnup ,.

SNF from conventional urania LWR assemblies has between 0.75 and 1% Pu as a result of

breeding from U238_.The Pu enrichment in the SNFfromdispositionedfuel is more nearly .... .

on the order of 1.8% for the reference fuel design. For example, the amount of Pu in a

single SNF assembly (for the reference design which uses 3.5% Pu enrichment), is 3.3 kg

of which 1.4 kg is Pu 239 and 1.2 kg is Pu 240. In disposition options where the initial Pu

enrichment is higher, the dispositioned SNF will contain more Pu. No criteria are available

relating the amount of Pu ir_ a single assembly to proliferation resistance, nor is it clear

whether there is a basis for such criteria. Given the small differences in the amounts of Pu

per assembly, it is judged that the proliferation resistance of SNF from the dispositioned

fuel and commercial fuel is comparable.

Conclusions:

Available (open literature) historical data base on SNF from commercial urania fuel

would appear to demonstrate that SNF possesses a very high level of proliferation

resistance with the implementation of appropriate safeguards. It is therefore concluded that

the proposed method of dispositioning Pu as SNF constitutes a verifiable and demonstrated

means of making this material proliferation resistant with existing technology.

2.2.3-2

2.3 QUALIFYING AND LICENSING MOX FUEL

2.3.1 REVIEW OF MOX FUEL LICENSABILITY

The considerations relative to lead assembly testing to qualify MOX fuel were examined in detail

during Phase 1B of this study and reported in Section 2.2 of Reference 1. During this phase of

this study, a more detailed examination was conducted with the following objectives:

(a) Complete a review of conformance to NRC requirements to license MOX fuel design inthe PDR

(b) Develop a program plan to test MOX fuel assemblies in existing research or industrial

reactors on an accelerated schedule consistent with existing NRC requirements, and

(c) Analyze U.S. Industry capability to support either BWR or PWR lead assembly testing.

Based On the evaluations givenin the Phase 1B report and further developed during Phase

1C, it is concluded that:

• A preliminary license for the use of MOX fuel could be granted based on the extensive

data already available for this fuel without any lead tests

• Limited lead tests, primarily in the form fuel of rods rather than fuel assembly

irradiations, might be required, with the specific objective of verifying the performance of MOX

fuel fabricated by the recommended process

• The lead test program should be used to validate the interface between the processes now

being developed for the destruction of the pit shape and its subsequent processing to fabricateMOX fuel

• The lead test program should be used to validate any new production techniques

including verification of procedures arising from automation/safeguards implementation

• All anticipated questions relative to final licensing could be resolved with either further

evaluations of the existing data base or from lead tests within 4 years of the program inception.

• Full scale MOX assembly tests are not needed for licensing. Limited full MOX assembly

tests might be conducted for confirmatory purposes.

PHASE 1B EVALUATIONS SUMMARY:

In Phase 1B it was concluded that although an extensive literature and data base are

2.3-1

available for Mixed (uranium-plutonium) Oxide (MOX) fuel properties and its performance in

LWRs, there are a number of reasons to undertake an early program of Lead Testing (LTA) in the

plutonium disposition project. The reasons included:

a. The need to develop and demonstrate a fuel fabrication process that complements the

proposed processes for destruction of the pit shape. A process of hydriding followed by

dehydriding has been proposed (Ref. 2) to destroy the pit shape. Currentlyexperimental

verification of this process is under way at LANL and LLNL.-Although additional work remains to

be done, it is clear that any proposed MOX fabrication process should complement the work

already being done in this area and still yield high quality MOX pellets.

b. Although urania-gadolinia pellets have been fabricated in large quantities in the past and

no difficulties are anticipated in fabricating MOX - gadolinia pellets, there is little data on

" fabrication of MOX mixed with the Gadolinia or other burnable poisons such as Erbium oxide or

Europia. There is a need to verify the fabrication parameters for this fuel.

c. There is a general need to reestablish the MOX fabrication technology in this country as

no large scale MOX fabrication has been carried out in more than 20 years. In this context, it is not

only necessary to reestablish the old technology which could probably be accomplished very

quickly but to define and incorporate safeguards and automation techniques to improve

productivity and lower worker exposure. The LTA program could serve as a vehicle to validating

any new production technique.

d. While nuclear analysis codes are readily available incorporating the full complement of

cross-section information needed for the design of a MOX core and the nuclear performance of

MOX fuel has been verified using partial core loads of this fuel in LWRs, additional benchmark

data for the nuclear performance of MOX fuel with burnable poison would help to better calibrate

the methods already in-place. Both self-shielding effects and poison burnout profile as a function

of radius within the fuel rod are considered important information that need to be benchmarked by

experiments.

e. It is desirable to obtain axial, radial power profiles and thermo-mechanical performance

data for full length rods from a MOX fuel assembly to provide confirmation of the code predictions

with experimental data.

2.3-2

A preliminary LTA (Lead Test Assembly) program was proposed with the following major

objectives: (1) to validate the fuel fabrication process, (2) to verify the thermal and mechanical

behavior of the MOX fuel thus fabricated, (3) to verify fuel rod nuclear performance and finally (4)

to verify the integral assembly performance. The plan proposed consisted of a three pronged

approach, first verification of fuel fabrication, second verification of individual fuel rod

mechanical, thermal and nuclear performance principally by conducting testsin research reactors

and finally verification of assembly performance through full scale MOX assembly testing. It was

pointed out that the last of these, full scale MOX assembly tests, was not considered essential for

validation and licensing of MOX fuel in PDR. This is because tests of isolated MOX assemblies in

existing reactors did not provide any additional data on overall system response as this - the overall

system response - will clearly be decided by the preponderance of urania assemblies in the core. In

addition, it was concluded that such full scale MOX assemblies will most likely be of a different

design so as not to perturb the already licensed envelope of the existing reactor core, in particular

the performance of adjacent urania assemblies. Therefore, tests of full scale MOX assemblies are

only of limited value.

During the present phase of this study, a more critical review of these aspects to MOX

qualification was undertaken. A step by step review of what it takes to license MOX fuel to NRC

requirements was conducted. The details of this review are presented in Appendix A while the

overall summary is included in this section. It is concluded that the most important aspect of

licensing the MOX fuel, considering that a vast MOX fuel data base already exists, is to fabricate

MOX fuel of sufficient quantity with the desired microstructure by a validated process and to show

its thermal-mechanical performance by rod tests in a research reactor. Based on this review, the

program plan presented during Phase 1B was updated. Finally, the industry capability to support

lead testing was evaluated.

The Standard Safety Analysis Report (SSAR) for the ABWR was submitted to the NRC in

1987. The NRC review is approaching completion. All major technical issues are resolved

(SECY-89-153 and SECY-90-016) and approved by the NRC on June 26, 1990. Final design

approval was expected but was rescheduled for 1994.

As presently conceived, and as evaluations to-date have shown, GE's ABWR will be used

for Pu disposition without involving any system level modifications whatever. The only change

relates to the use of MOX fuel instead of the standard urania fuel. New fuel licensing is carried out

by a process of documenting the compliance with the criteria of Amendment 22, a process

2.3-3

described below in more detail. Line-by-line compliance status of the Amendment 22 criteria is

included in Appendix A.

In developing the ABWR, two bases were used for the analysis -- design basis and

licensing basis. For the design basis of the Nuclear Boiler, GE chose the most conservative core

design that was known at the time, including the possibility of using mixed-oxide fuel in the

future. This was designated the Core Z designused for the plant design development. In the

licensing basis, reported in the SSAR for limiting transients, GE used a reference fuel design,

typical of today's UO2 fuel offerings, called, Core A. Thus, asalternative mixed-oxide fuel

designs used in this present study were developed,-GE already had a guide as to which transients

and accidents would be the most limiting, and whether there might be problems accommodating the

designs.

A report of the description of the fuel licensing acceptance criteria for the GE9 fuel design

is specified by Amendment 22 of the GESTAR II document (General Electric Standard Application

for Reactor Fuel, NEDE-24011-P-A). The amendment contains the basis for generic compliance

of the GE9 fuel design with those criteria. The reference fuel design for Pu disposition in fact uses

this standard GE9 fuel design with the exception that the fuel will be MOX fuel instead of thestandard urania fuel.

The proposed method of disposing Pu by the use of its conversion to MOX fuel and its use

in an ABWR entails no change whatever to the reactor system design except for the use MOX fuel

in the standard GE9 fuel design. Therefore, a detailed evaluation of Amendment 22 with respect to

this change should, as a start, establish the licensability of the ABWR system and the use of MOX

fuel. In this evaluation, two different approaches are possible:

a. To describe a specific MOX fuel design and establish the licensability of this particular

fuel design, or

b. To establish the licensability of MOX fuel in a generic manner, similar to the generic

licensing of Urania fuel under the GE9 fuel design, that is, to establish the licensability of MOX

fuel over an umbrella design range.

It should be noted that in either case, the satisfaction of these criteria will require cycle-

unique analyses which must be performed after the core loading for that'cycle has been specified.

For these cases, the generic information contained in this section will be supplemented by plant

cycle-unique information and analytical results. In general, this cycle-unique information will be

documented in a separate cycle-unique reload licensing report for each reload.

As reported in Phase 1B, if the intent is to obtain approval of MOX fuel for the government

owned PDR for operation on a government reservation, rather than to obtain approval for

2.3 -4

commercial operation of a BWR with MOX fuel, it is likely that the fuel safety review will be

conducted as part of the SSAR review, and the GESTAR process will not be used formally. Even

if the PDR is owned and operated by an independent power producer with the Government

providing the fuel, it appears likely that a special MOX fuel licensing review will be conducted

since present government policies do not envision private licensing of MOX fuel. However, the

technical elements are expected to be identical to those of the GESTAR Amendment 22 compliance

review process. Therefore evaluation of the GESTAR steps for licensing MOX fuel is relevant tothe PDR.

When the NRC is notified that the generic analyses for the new MOX fuel design are

completed and all criteria are satisfied, the fuel design is "licensed". No prior NRC review and

acceptance is required. If a specific criteria is not met, prior NRC review and acceptance is

required; however, the review is limited to the area of noncompliance. The reference UO2 fuel

design for the ABWR meets the acceptance criteria of Amendment 22 as documented in Appendix

4D of the ABWR SSAR. No further NRC review is required if the Combined Construction and

Operating License applicant utilizes this fuel design. It is the current NRC position that any change

to the reference fuel design for initial core application (e.g.. MOX designs) will require a change to

the SSAR and NRC review. However, if the NRC has been previously notified that the MOX

design meets the Amendment 22 criteria, this review should be only a formality.

An assessment has been completed to determine the impact on the licensing process for an

ABWR assembly with MOX substituted for the UO2 fuel. It is concluded that the MOX fuel

design will have to be treated as new fuel and compliance with Amendment 22 acceptance criteria

has to be demonstrated. There are three areas where the use of MOX fuel is expected to have an

impact on the licensing process. These are:

1. Amendment 22 requires the use of NRC approved analytical models and analytical

procedures. This will require prior NRC review of the analytical methods if the current NRC

Safety Evaluations for these methods do not address MOX applications.

This area was treated in some detail in the Phase 1B report. No changes to any of the

thermal-hydraulic or structural models are needed. While no changes to the basic methods already

in place for nuclear analysis are needed, calibration of the diffusion/transport models used in

nuclear analysis would be needed. This calibration will be performed by comparison with the

results of the _ _re exact Monte-Carlo solutions. The results of preliminary studies in this area

were reported in Phase lB. Additional calibrations require more detailed analyses which will be

2.3-5

performed in Phase 2. Once the calibrations are in place, utmost small changes might be indicatedto the Pu enrichment.

2. Amendment 22 requires that new design features be included in Lead Use Assemblies

(LUAs). If MOX is considered a new "feature" LUAs will be required before MOX is licensed.

As presently designed, MOX fuel assemblies are not considered to presentany new

"feature". The assembly thermal-hydraulic and mechanical response is expected to be identical in

every respect to the already licensed GE9 bundle design. The effect of using MOX fuel is internal

to the pin and the technology elements that need to be tested relate to the behavior of the fuel rather

than the bundle. This area is clarified in more detail in Appendix A. In summary, it is concluded

that while MOX pin tests are needed with a view to confirming the behavior of the fuel fabricated

by a representative process and which has undergone the necessary quality assurance steps, LUA

tests are not considered necessary.

3. Amendment 22 includes a general criteria that encompasses "new-fuel-related licensing

issues identified by the NRC." For MOX fuel this could result in the NRC request for new criteria

to address the concern(s).

It is this change, the potential changes to the behavior of the fuel, which might result in

new limits, that needs to be assessed in detail. This is the area that is covered in detail in this

section. It should be pointed out that typically LWR fuel at discharge contains up to 1% Pu that is

bred from the uranium during the course of irradiation. The reference fuel designs recommended

by GE for Pu disposition generally are small extrapolations in that the average Pu enrichment is

kept below 5% in all cases and for the reference case is below 3.5%. The wealth of data available

from MOX fuel irradiations fabricated from reprocessed fuel as well as the data generated in DOE

sponsored programs on MOX fuel containing up to 20% Pu for Liquid Metal Reactors in the past,

provide the necessary foundation on which we can confidently predict the behavior of MOX fuel in

the ABWR. This is supplanted with fuel pin tests to be carried out under the proposed LTA

program.

Therefore, from a licensing perspective, the two areas that need to be addressed are, first,

the calibration of nuclear methods for MOX applications and second, the effect of changes in

physio-chemical properties of the fuel in going from urania to MOX and its impact on fuel thermo-

mechanical behavior. A summary of the evaluations in these two areas is given below:

MOX FUEL ROD NUCLEAR PERFORMANCE: CODE CALIBRATION

2.3 -6

Confirmation of the core nuclear performance might require further calibration of the codes

now in use for urania fuel. The principal objective will be to conduct selected rod and assembly

tests to provide the required data for calibration of the diffusion/transport nuclear codes, in

particular for the rate of poison burnout as a function of radius in a MOX fueled rod. In addition,

it is expected that extensive cross-calibration will be done using Monte-Carlo codes.

MOX FUEL ROD THERMAL-MECHANICAL PERFORMANCE ,.

Fuel rod thermal-mechanical performance could be expected to be affected by the change

from urania to MOX fuel. A technical review of the available urania-plutonia mixed oxide fuel

properties and performance information has been conducted to qualitatively assess differences,

with respect to urania fuel, that may affect fuel rod thermal-mechanical performance relative to

design and licensing criteria. The results of this review given below indicates that for the range of

Pu enrichments considered in this study, the thermal-mechanical response of the fuel is predictable

using the existing data base and the fuel is expected to perform as well as the standard urania fuel,

The specific fuel properties and performance characteristics investigated include:

Theoretical density

Elastic modulus

Creep

Thermal expansion

Thermal conductivity

Enthalpy

Melting temperature

Radial power distribution

Fission gas release

Theoretical Density:

Plutonia lattice parameter measurements exist to enable a quantification of the urania-

plutonia theoretical density as a function of the plutonia concentration. The fuel theoretical density

(gm/cc) increases with increasing plutonia concentration; the theoretical density of 10 w/o PuO2 -

UO2 is - 0.5% greater than the theoretical density of urania. The effect of this difference on fuel

rod thermal-mechanical performance is that, for the same fuel exposure accumulation and fraction

of theoretical density, a slightly greater number of fissions would occur per unit volume of fuel

2.3-7

thereby leading to slightly greater fuel irradiation swelling and gaseous fission product inventory.

The difference, however, is relatively minor.

Elastic Modulus:

Measurements of the elastic modulus of plutonia indicate that the elastic modulus of urania-

plutonia fuel increases with increasing plutonia concentration; the elastic modulus of 10 w/0 PuO2

- UO2 is ~ 1.5% greater than the elastic modulus of urania. ,This difference.results in a slightly

stiffer pellet, and correspondingly higher stresses for a given elastic strain. This difference is

relatively minor and overshadowed by the difference in fuel material creep behavior describedbelow.

Creep:

Measurements of the temperature-dependent creep behavior of urania-plutonia

compositions indicate a significant increase in the fuel creep rate relative to UO2; the creep rate of

10 w/o PuO2 - UO2 at representative operating temperatures and stresses is - 3 times higher than

that of Urania under the same conditions. This difference indicates a more compliant fuel pellet

and correspondingly lower fuel rod cladding stresses for the same loading conditions. This

difference also indicates a lower pellet-cladding interfacial pressure and correspondingly, lower

pellet-cladding thermal conductance and higher fuel temperatures for the same loading conditions.

Thermal Expansion:

Measurements have been performed to determine the isothermal thermal expansion

coefficient of plutonia. These measurement results indicate decreased fuel thermal expansion with

increasing plutonia concentration; the thermal strain at 2500F for 10 w/o PuO2 - UO2 is -- 95% of

that for urania. The effect of this difference on fuel rod thermal-mechanical performance is a

slightly reduced imposed strain on the fuel rod cladding by the fuel pellet (for the same fuel

temperature condition) and correspondingly reduced cladding stresses under certain loading

conditions (such as a rapid power increase).

Thermal Conductivity:

Measurements of the thermal conductivity of plutonia and urania-plutonia compositions

have been performed by drop calorimetry methods. The measurement results consistently show a

slightly decreasing fuel thermal conductivity with increasing plutonia concentration, the thermal

conductivity of 10 w/o PuO2 - UO2 at typical operating temperatures is ~ 97% of that of urania.

The effect of a lower fuel thermal conductivity is to increase fuel temperatures, and

2.3-8

correspondingly fuel thermal expansion and fission gas release, for the same operating powerlevel.

Enthalpy:

Enthalpy measurements of plutonia and urania-plutonia compositions have been performed

by drop calorimetry methods. The measurement results indicate a slightly higherenthalpy than

urania at lower temperatures and a slightly lower enthalpy at higher fuel temperatures; the enthalpy

of 10 w/0 PuO2- UO2 is ~ 1% higher than that for urania at 816C (1500 F) and ~ 1% lower than

that for urania at 1371C (2500F). This difference is negligible.

Melting Temperature:

Extensive measurements using the thermal arrest technique have been performed by GE to

determine the solidus and liquidus boundaries of urania-plutonia over the entire composition range

..... (0-100% PuO2). The measurement results indicate a decreasing melting temperature (solidus) with

increasing plutonia concentration; the melting temperature of 10 w/0 PuO2 - UO2 is ~ 60C

lower than that for urania. One US licensing constraint is that fuel temperatures not exceed the

melting temperature during normal steady-state operation, including anticipated occurrences.

Therefore, this lower fuel melting temperature will reduce, to a small extent, the operational

capability of the fuel.

Radial Power Distribution:

For urania fuel, the radial power distribution across the fuel pellet is relatively fiat at the

start of irradiation. With continued irradiation, the progressive buildup of plutonium near the fuel

pellet outer surface causes a significant peaking in the radial power distribution near the fuel

surface, with a corresponding depression of the power in the pellet interior. The effect of this

time-varying change in the radial power distribution is to reduce the pellet centerline temperature

for the same power level and pellet surface temperature. For urania-plutonia fuel, the presence of

plutonia near the pellet surface causes an increased peaking (relative to urania) over a greater region

toward the pellet surface even at the start of irradiation. This increasingly surface-peaked radial

power distribution persists throughout lifetime for urania-plutonia fuel. The effect of this

difference is slightly reduced fuel central temperatures, relative to urania fuel, for the same

operating power level and fuel surface temperature.

Fiss;on Gas Release:

The release of gaseous fission products from the fuel to the fuel rod void space produces

two primary effects:

2.3-9

I

(1) The thermal conductivity of the gaseous fission products is approximately an order

of magnitude lower than the helium filler gas introduced during the fabrication of the fuel rod.

Therefore, the release of gaseous fission products from the fuel pellets to the void space reduces

the gas mixture thermal conductivity, reduces the conductance between the fuel pellet and the

cladding, and increases fuel temperatures. The fission gas release mechanism is fuel temperature

dependent, and therefore, this fueltemperature increase produces additional fission gas release,

(2) The release of gaseous fission productsto the fuel rod void space increases the fuel

rod internal pressure. The maximum fuel rod internal pressure is limited by US licensing

constraints. Therefore, excessive fission gas release and fuel rod internal pressure can limit the

operating conditions or mechanical design of the fuel.

The available information indicates that the fission gas release behavior of urania-plutonia

" " fuel is significantly affected by the as-fabricated fuel microstructure. For example, significant open

porosity (fuel pellet pores in direct contact with the pellet surface) increases the rate of fission gas

release. In addition, inhomogeneous microstructure resulting in localized islands of pure PuO2

and/or UO2 also increases the rate of fission gas release. The preferred microstructure includes

standard grain sizes (8 - 10l.tm), low open porosity (<0.5% TD), and 100% homogeneous

(U,Pu)O2 solid solution. With this preferred fuel microstructure, no difference in the fundamental

fission gas release processes and behavior are expected relative to urania fuel.

US fuel licensing criteria can be broadly categorized as either (1) thermal performance

limits (e.g., fuel melting temperature limit), or (2) mechanical performance limits (e.g., fuel rod

cladding stress and strain limits). The urania-plutonia fuel properties and performance assessment

indicates that, relative to urania fuel, fuel thermal performance is somewhat less favorable and fuel

mechanical performance is somewhat more favorable. These differences, are, manageable and can

be accommodated either by the thermal-mechanical design of the fuel rod or specification of

operating constraints. The performance area that can be influenced to minimize performance,

design and licensing differences is fuel pellet fission gas release; application of known preferred

microstructural features will eliminate the potentially significant difference.

Based on the above evaluations, it is concluded that a preliminary license could be granted

for use of MOX fuel without additional testing. Specific issues could be raised by the licensing or

reviewing agency which could be addressed in the LTA program.

2.3-10

References:

1. "Study of Pu Consumption in Advanced Light Water Reactors, Compilation of Phase 1B

Task Reports," GE Nuclear Energy, RFP DE-ACO3-93SF19681, September 15, 1993.

2. Haschke, et. al.,"A Hydrogen Recycle Process for Plutonium Recovery," LA-12086-MS,

Lawrence Livermore National Laboratory, 1991.

2.3-10A

2.3.2 PROGRAM PLAN FOR LEAD FUEL TESTING

The Lead Test Assembly program outlined below was developed by GE during Phase IB of

this study. In developing this program, input was received from all disciplines within GE as well

as from national laboratory experts at LANL, LLNL and WHC. The program envisions obtaining

all the necessary confirmatory data for fuel fabrication within 12 months of program initiation,

verification of individual rod behavior within 36 months and final licensing of MOX core through

calibration/validation of nuclear analysis within 48 months.

These projections are based on pursuing an aggressive schedule, not limited by funding

availability.

A. Objectives"

The objectives ol. is program are:

Fuel Fabrication ValidationFuel Mechanical/Chemical Performance Validation including Fuel Properties Data-Base

Generation if RequiredFuel (Rod) Nuclear Performance Benchmark Data Generation

The objectives under each of these categories are described in more detail below:

Fuel Fabrication Valid_tiQn

• Develop MOX Processes to Complement Pit Processing

As a first step in safeguarding the weapons plutonium, the pit which contains the plutonium

in a weapons has to be removed and its shape destroyed. Although a number of techniques are

available, research in this area has focused on a "chemical processing,' means as a clean, safe and

economical means of pit shape destruction. This process consists of hydriding and dehydriding the

plutonium. In such a process, the pit shape falls apart leaving a powdery plutonium as the final

state. It is therefore useful to consider MOX processing steps which complement this process. If

the fission process is not chosen as the route to plutonium disposition, the Pu powder from this

process would presumably be melted or converted to an oxide and stored in a desired unclassified

shape.

2.3-11

The plutonium obtained from the hydride-dehydride process could be directly converted,

under a controlled oxygen partial pressure, to produce plutonium oxide. This could form the

starting stock for the MOX process with urania and gadolinia powders to be supplied by a

commercial vendor such as _3E. However, the activity of the plutonia powder and the particle size

distribution thereof, may not be compatible with what is required for MOX fuel fabrication.

Therefore, further studies on the milling of the plutonia powder and blending.with the urania have

to be conducted under controlled conditions to develop an acceptablemixed MOX powder for

sintering. In this regard, the minimum acceptable plutonia particle size in a MOX pellet is already

available based on earlier studies. Initial fabrication development will therefore concern itself with

the ability to meet the final fuel specifications and quality requirements in a number of respects

including density, grain size and Pu distribution specifications. The urania required for this

fabrication evaluations could be provided by GE or the laboratory could find its own source for

this supply. Specifications for the MOX fuel pellet will be provided by GE as well as the post-

.......: _ _sintedng ex:amination requirements.

• Demonstrate Fabricability of MOX with Gd

Although an extensive MOX fuel fabrication .data base exists, there is relatively little

experience in the way of fabricating MOX fuel with Gadolinia. GE has extensive experience in

fabricating Urania fuel with Gadolinia burnable poison and will provide the initial sintering

parameters which might be expected to work with MOX fuel. A major parameter of interest is the

oxygen overpressure during sintering to produce acceptable final chemistry. It will be necessary to

inspect and analyze fuel sintered over a range of oxygen overpressures to ascertain the correct

sintering atmosphere to be used in production. It is also proposed that a range of Gd-Pu contents

be examined under this evaluation. A maximum of 10% Gd (by weight fraction) is suggested as

the upper limit for the investigation. It is expected that the Gadolinia powder will be provided by

GE together with the specifications for this powder. Post-sintering examinations will be carded out

by the laboratory based on examination specifications to be provided by GE.

• Effect of Impurities on Processing and Sintering

Americium and other metallic impurities are expected to be present in minor or trace

quantities in the initial plutonium feed. There is a need to reduce their presence as much as

possible. High levels of Americium will tend to increase worker exposure. The processes to reduce

or eliminate them from the initial plutonium feed material are well established but need to integrated

2.3-12

with the overall process flow. It has been suggested that by building the fuel fabrication factory to

allow complete remote automated handling, the level of Americium that can be present could be

raised significantly. This does not however eliminate subsequent problems in handling the fuel

bundle transport to the reactor and receiving inspection where high Americium levds lead to higher

worker exposure. The proposed concept therefore aims at reducing these impurities to a low

enough level so they do not pose any problems downstream. In this regard, some fabrication

development and post-sintering examinations are required. In addition, it is also known that

Americium preferentially evaporates during sintering. Data from the initial runs will beuseful in

acc0tinting for possible worker exposure and designingpreventive systems to trap the evaporatedAmericium.

• Validation of Fuel Pellet Quality

• GE wilt be providing the fuel pellet specifications including those for density, grain size, impurity

limits and fuel chemistry. Post-sintering examinations should be conducted on sufficient pellets to

Validate the overall fabrication process and individual process steps to aid in fine tuning the MOX

Fuel Factory Requirements, Equipment Specifications and Process Parameters. Although sintering

parameters for MOX fuel are well established, a limited number of additional studies should be

conducted to optimize the process, particularly with a view to reducing the waste stream and scrap

recycle fraction which are the major contributors to worker exposure.

Fuel Mechanical/Chemical Perf0rmanc¢ and F0¢I Pro_nertie_:

• Mechanical/Chemical Performance

The objectives here are simply to verify the pin mechanical/chemical performance of the rod

and verify its integrity to goal exposures. Fuel rods will be fabricated and irradiated in research

reactors closely simulating the rod parameters expected in the disposition reactor. Post-irradiation

examinations will include among others: gathering fission gas release data and pin dimensional

(strain) measurements as a function of fabrication/operating variables; checking the migration of

specific fission products through gamma scanning;' fuel length change measurements; fuel-cladding

interface examinations for chemical compatibilit.y. The fabrication variables will include Pu-Gd

fraction, fuel power density and the range of allowable fuel pellet physical specifications.

• Fuel Properties

2.3-13

Properties of MOX fuel are readily available from previous DOE programs including the

Liquid Metal Fast Reactor Development Program. However, limited additional data are needed,

particularly for MOX fuel with Gadolinia poison. Material property correlation models for oxide

fuel containing small fractions of rare earth materials (of which Gd is one) have been developed at

GE, nevertheless, experimental confirmation of these data might be requested in licensing review.

Two Specific properties of major interest are: fuel thermal conductivity, and fuel solids, liquids

temperatures. Pellet Oxidation tests inoxygenated water might also be requested in order to answer

any questions relative tothe performance of a breached pin for the limited duration it remains in the

reactor prior to removal.

Fuel (Rod) No¢lear Pcrfgrmance Benchmark Data

' -The primary objective here is to confirm the nature of Gd bumup as a function of the radius

within an individual rod. Different Pu-Gd compositions could be examined. Rods can be irradiated

individually or in a cluster, in an experimental reactor such as the ATR. GE will provide the test

specifications and post-irradiation examinations and conduct the associated nuclear analysis. A

range of fission densities should be considered in the experiment. The results will provide the

benchmark data for confirmation/validation of nuclear codes and refine the nuclear design.

B. Preliminary Test Plan

Based on the foregoing objectives, the following test plan is proposed:

Fuel Fabrication Validation Tests

a. MOX Fuel will be fabricated by a DOE designated laboratory based on mechanically

mixed process, with Pu oxide from Pu feed stock prepared from weapons grade Pu or comparable

chemistry. The process specifications from the Pu Oxide phase to MOX fuel will be arrived at

jointly between GE and the fuel fabrication laboratory.The final specific_itions for the MOX pellet

will be provided by GE. The final specifications for QA for the fuel will be specified by GE. GE

will provide the initial input for sintering Gd bearing MOX fuel. The fabrication procedures will

also attempt to verify any key elements related to automated production/safeguards implementation

that may be required. For example, this could consist of active interrogation of the Pu content and

inventory logging in on-line computer systems.

2.3-14

b. Range of Fuel Pellet Parameters:

Different Pu-Gd compositions (Gd from 0 to 10% and Pu from 2 to 20%)

Nominal Pellet density: 96.5%

Nominal Grain Size: 10 microns

Specific Pu-Gd combinations to be specified by GE in Phase 2.

c. Post-fabrication tests for:

Optical Metallography (etched and unetched)

Fuel chemistry (O:M Ratio, impurities)

Pu homogeneity

Fuel density, description of porosity (by metallography)

Surface Roughness Post-fabrication examinations to be specified by GE; these tests

will be conducted on various batches to describe any effect of process variables on

final fuel pellet characteristics

d. Fuel properties tests:

Thermal conductivity (by laser flash technique)

Thermal Arrest Studies

(Above tests for various Pu-Gd compositions)

Fuel Creep for selected compositions

e. QA of reference fabrication process pellets

GE to specify QA requirements; laboratory will conduct the QA audit of the fabricated

The results of this series will be documented by the.laboratory with particular emphasis on

process verification for production of Gd bearing and non-Gd bearing MOX fuel.

In-Reactor Rod Tests

The in-reactor rod tests are aimed at meeting the nuclear benchmark data generation and

thermo-mechanical performance confirmation objectives. The rods will be assembled by the

2.3-15

laboratory and shipped to the test site. The test site will be responsible for the in-coming

inspection, thermal-hydraulic design of the test, for obtaining preliminary and final approvals for

the test and for the safety analysis based on any required input from GE and the fabrication site._

The detailed test specifications will be provided by GE. Either individual rods or a cluster of rods

will be irradiated in a research reactor. The power level and rod physical parameters (diameter and

length) will be provided by GE.

As currently envisioned, these tests could be individual capsules or a cluster of capsules.

None of the tests in this series have any active monitoring of either the temperature or for any other

parameter. Indirect confirmation of the temperature could be obtained by incorporating TEDs

(Thermal Expansion Devices) which have been used in the past. To the extent these tests are

entirely uninstrumented and passive, the cost of these tests should be low and an aggressive

schedule could be pursued. Data will be obtained as a function of exposure by removing the

capsules containing the rods at regular intervals, to be specified by GE.

Post-irradiation examination requirements will be specified by GE and will include, as

noted earlier, gamma scans, fuel length changes, rod profilometry, limited fuel pellet

metallography, fission gas collection, and limited microprobe examinations of the fuel-cladding

interface. These examinations will be conducted at a suitable facility to be designated by DOE. GE

will conduct Gd radial profile measurements in its VNC facility from shipment ef samples to be

made from the designated post-irradiation examination facility.

The details of the test matrix will be worked out after meetings between the interfacing

organizations. As presently envisioned, the test matrix will include the following parameters:

Fuel Characteristics:

MOX ROd without Gd for Comparison with Gd with Pu as variable

Pu-Gd Composition (Up to 10% Gd)

Fuel density, grain size within allowable range

Operating Conditions:

Exposure

Power Level

2.3-16

Because these tests are designed to be uninstrumented, fuel behavior as a function of

exposure will be obtained by removing the capsules after specified exposure levels rather than

monitoring pin behavior as a function of time while in-reactor. Individual rods or rod clusters

(depending upon the test cavity size) will be irradiated in a reactor such as ATR.

Assembly Tests

A limited number of full length MOX assembly tests are recommended, primarily to

confirm the performance of full length rods. These tests are not however anticipated to provide

any additional data for licensing. In conducting full assembly tests, it should be recognized that the

MOX fueled assembly will be placed in a sea of urania fueled bundles and the nuclear performance

of the reactor system will be controlled by the urania fuel rather than by the isolated MOX fuel

assemblies. A preliminary assessment has indicated that with the older reactors only partial MOX

• 'core loads are possible. In point of fact, the MOX fuel test assembly has to be designed to

minimize its impact on adjacent urania fueled assemblies and to operate within the limits of the

criteria for which the plant was originally designed. A number of design features in the ABWR, as

pointed out in the Phase IA report, allows it to accept a full core of MOX fuel while this is unlikely

to be the case with older LWRs. In addition to these factors, it should be borne in mind that if the

test were to be conducted in a conmaercial facility, it might be necessary to match the cycle length

of the specific core and an independent specification of the test fuel exposure may not be possible.

For these reasons, MOX fuel assembly tests in a reactor are not considered necessary for

licensability. Rather, it will be the confirmation of individual rod tests and their use as benchmarks

in the nuclear analysis codes that would be used for proving the nuclear and thermo-mechanical

performance of MOX fuel in an LWR.

If full scale MOX assembly tests are requested, two types of LTA assemblies are under

consideration and these will be examined in more detail at a later date. An assembly consisting of

all MOX rods could conceivably be inserted in the outer periphery and data could be obtained on an

all MOX bundle. A second type of assembly will utilize the so called "island" design whereby the

MOX rods are placed in the interior of the assembly and are surrounded by urania rods, so that the

LTA assembly's effect on adjacent fuel assemblies is minimal.

Full assembly tests could be used to study the effect of different Gd/Pu enrichments and

provide data on axial and radial power shaping in the assembly. Data will be obtained on rod

profilometry, length changes and fission product migration by gamma scanning, on a cycle-by-

cycle basis. GE will be providing the test specifications, MOX fuel will be fabricated at a

2.3-17

i

designated laboratory and assembled and shipped to the test site. Where urania rods are needed,

these could be provided by GE.

Facilities for examining full length rods are expected to be available at GE-VNC. A number

of interface issues relative to shipping, receiving and disposal have to be worked out, however,

these can be tackled once it has been determined (by the licenser/reviewer) that such full scale

MOX assembly tests are needed and agreement has been reached for inserting MOX bundles into

an operating BWR.

C. Interfaces

In order to carry out this program, a number of interfaces have to be established and the

activities coordinated. A PDR-LTA Interface Control Board may need to be established with

specific responsibilities and assignments and charged with the conduct of the program. The type of

interface arrangements that are suitable for implementing the lead test program was outlined in the

Phase 1B report.

D. Cost:

Detailed cost estimates can be generated after an initial licensing review to more clearly

define the areas where additional data are deemed necessary. It is anticipated that approximately

50 to 60 capsule tests might be needed. Cost of the program for fuel fabrication development and

for use of the test reactors at the interfacing sites will have to provided by the site chosen for these

activities. As a rough order of magnitude, it is estimated that the rod tests would cost up to $2

million at the te_t site including the post-irradiation examinations. The cost for fuel fabrication

activities will depend upon whether a facility has to be dedicated to this task or whether facilities

(glove box lines) where Pu is being handled is already available. Initial rough cost estimates are

approximately $2 to 2.5 million per year for a period of 3 years. The fuel properties data generation

tests are expected to cost approximately $0.5 million per year for a period of 3 years. The GE

resources needed for this program is estimated at 6 full time engineers and 2 senior professionals

for integrating all the external and internal interfaces.

E. Schedule:

Details of a proposed schedule were presented in the Phase 1B report. At that time it was

assumed that the lead testing program would be initiated in FY94. The previously suggested

schedule has been reevaluated and no changes are suggested for the duration for the various

2.3-18

activities. The following durations for the major milestones had been identified during Phase 1B:

(all dates after program start)

1.Produce initial MOX test fuel: 12 months

2. Complete Fuel Fabrication Development: 24 months

3. Complete Fuel Properties Measurements, if required. 24 months4. Initiate Rod Test irradiations: 12 months

5. Complete Rod Testing and Issue Validation Reports: 48 months

6. Initiate Assembly Irradiations, if required: 30 months

7. Complete Assembly Irradiations and Issue Confirmatory Reports: 72 months

Based on the evaluations presented, it is once again concluded that preliminary licensing for

MOX core utilization could be granted based on already available data and any remaining issues

will be answered with the completion of milestone 5 above or in 48 months after program

initiation. The full scale assembly test data are not needed for licensing but may be useful to refine

"the power shape of down-stream cores.

2.3-19

2.3.3 US INFRASTRUCTURE FOR LEAD FUEL TESTING

The infrastructure for implementing the disposition project, discussed in Section 4 of this

report, also covers several aspects related to the lead testing program. In summary, an extensive

infrastructure that is ready and capable is available to support the LTA program, specifically in the

following areas:

a. To define the Pu interface between the LTA fuel fabrication and pit material (LLNL,

LANL)

b.To identifytheprocessrequirementsandfabricateinitialbatchesofMOX (LANL)

c.To furtherdefinetheprocess,incorporaterequiredautomation/safeguardsverifications

(GF_ANL/LLNL)

d.To shipMOX fuelrodstoresearchreactorsites(thishasbeendemonstratedinother

DOE programs)

e.Testsitestoconductrodtests:Reactorsareavailable(ATR,HIFR,others)

f. Post-Irradiation Examination Facilities: (WHC, INEL, GE-Vallecitos, others)

It is only in conducting full scale MOX assembly tests that a suitable reactor site and the

necessary interfaces have not yet been firmly identified. Inquiries have been made with several

BWR vendors and it is clear that isolated full scale MOX assemblies could be irradiated in existing

reactors. Initial inquiries have also indicated that a similar situation exists for PWRs, in that there

are at least some reactors which would accept MOX fuel assembly tests. A number of interface

responsibilities have to be worked out including those for licensing the test assembly, hardware

procurement, exposure limits, post-irradiation examinations and eventual disposal. However, firm

commitments could not be obtained because of the highly preliminary nature of this project at this

time. Before proceeding further with full scale MOX assembly tests, it will be necessary to obtain

from the licenser/reviewer the specific objectives to full scale MOX assembly tests since partial

core loads of MOX fuel have already been irradiated to significant exposures in both BWRs and

PWRs, as reponed in the final Phase 1A report.

Further discussions in this regard have also indicated that with the potential for private

funding for the reactor system by utility/IPP groups, these groups will take the responsibility of

identifying the reactor to conduct the MOX assembly irradiation tests and working out any interface

responsibilities. An alternative, if the disposition were to be entirely funded by the US

government, would be to use reactors which were built with government funding.

2.3-20

2.3.4 European Infrastructure for MOX Testing

In general terms, the performance of a new fuel type must be proven by irradiation in a

commercial reactor. This is requirednot only to demonstratethat the fuel behaves satisfactorily,

but also to allow the validation of models for the processes which occur during irradiation.

Performance mustbe proven not only for steady-stateoperation but for a varietyof frequent faultconditions. However, since commercialreactor fuel contains no instrumentation,no data can be

produced. Data can be obtained in two ways: from post-irradiation examination (eitherat

poolside or in hot-cells), or by conducting irradiationexperiments on instrumentedfuel in .a test

reactor. Commercial reactor irradiation is also relatively benign, whereas the validation of

licensing methodologies must cover more sever operation. Specialized test reactor experiments,

e.g., PCI failuretests, mustbe mountedto addressthese issues.

'- ......... °--These general,principlesareapplicableto the testing of MOX fuel containinghigh grade, weapons

plutonium. Clearly,however, it would be inappropriateto regardsuch a material as a completely

new fuel type, since it is closely relatedto conventionalMOX fuel with which there is reasonablyextensiveexperience. Nevertheless, claimingthe priorexperience with conventionalMOX fuel as

relevant is only reasonable if, (1) design of the new fuel rods does not depart significantlyfrom

the base of experience with conventionalMOX fuel (2) manufacturingprocesses used to produce

the new fuel have a proven track record for producing conventional MOX fuel (3) it can be

demonstratedthat the differencesin plutoniumisotopic composition between the new fuel and (4)

conventional MOX (as well as any other differences) do not significantly change the fuel's

performance.

The second of these points is very important. It has always been recognizedthat the performance

of MOX fuel is much moremanufacturingroute specific than is the case with standardurania fuel.

Fuel microstructurein general, and plutoniumhomogeneity in particular, are found to be sensitive

to the method of plutonia/urania blending and pellet sintering, and other aspects of the

manufacturingprocess. Achieving high plutonium homogeneity-has long been recognized as one

of the key factors in producinggood MOX fuel. With high-grade,weaponsplutonium, this factorwillbe even moresignificant.

It mustbe assumedthatthe new fuel willbe manufacturedusing processes that have been proven

to produce highly homogeneous MOX fuel, and that the fuel rod and assembly design will not

departappreciablyfromproven experience. Given this, the qualificationprogram can be confined

in scope to studying the differencesbetween the new fuel and conventional MOX fuel. The

expected differences arise from the different isotopic composition. Weapons plutonium will

2.3.4-1

consist of typically 95 % fissile isotopes, compared with perhaps 65 % for material from a

reprocessing plant. The absence of the absorbing Pu-240 will make weapons MOX appear, from

a neutronics standpoint, closer to standard enriched uranium fuel than normal MOX fuel. For a

given lifetime average reactivity, weapons MOX is likely to give higher reactivites at start-of-life

and (because of less Pu-241 build-up) lower reactivities at end-of-life compared with conventional

MOX. This will help to reduce the problem of high late-in.life powers than can lead to high clad

corrosion and fission gas release in conventional MOX. The disadvantage is that rod power

peaking factors will be higher and burnable poison loadings (such as gadolinia doping) would

need to be increased. The higher fissile fraction also leads to the greater sensitivity to plutonium

homogeneity discussed above.

2.3.4-2

2.3.4.1 Single Rod Tests

Should single rod tests be required, either in advance of or in parallel with demonstration

irradiationsin a commercialreactor, then the Halden reactor is available as the ideal vehicle to

performthese tests. This is a smallBWR run by an internationallyfunded group of which both

BNFL and GE aremembers. The group have considerableexperience in runningirradiationtests

on MOX as well as uraniafuel. Theirparticularstrengthis the development of rig designs and in-

pile instrumentationwhich can elucidate the real processes occurringduring irradiation. Several

of their most recent developments have been focused on obtaining data on high burnup, fuel

without the needfor the long dwell times that are normallyrequired.

Comparative irradiations (Urania versus conventional MOX versus weapons MOX) of a small

number of rods could be mounted quickly. With appropriate in-pile instrumentation, data on

......•"_hermal-performance, dimensional stability and fission product release could be generated.Performance data would become available as soon as the test was loaded. Such a test could be

giving important insights into the behavior of the new fuel many years before a commercial

reactorirradiationandPIE programcould cometo fruition.

2.3.4-3

2.3.4.2 Commercial BWR Irradiation

The key step in qualifying a new fuel type is to conduct the irradiation of one or more

demonstration assemblies under prototypic commercial reactor conditions. However, since such

an irradiation does not in itself produce any information with which to draw a conclusion, it must

be followed by a program of post-irradiation examination (PIE).i

It is first necessary to identify a suitable commercial BWR in which a trial irradiation could be

undertaken. There are currently 23 commercial BWRs operating in Western Europe. Their

capability with regard to irradiation of MOX fuel is summarized below:

Reactor Utility MOX Lieense Status

Oikiluoto 1 TVO License not applied forOikiluoto 2 TVO License not applied forSwedenBarseback ! Sydkraf_ License not applied forBarseback 2 Sydkrah License not applied forForsmark I FKA License not applied forForsmark 2 FKA License not applied forForsmark 3 FKA License not applied for

•Oskarshanm ! OKG License not applied forOskarshamn 2 OKG License not applied forOskarshanm 3 OKG License not applied for

Ringhals 1 SSPB License not applied for

Brunsbuttel KKB Submitted application in 1986Gundremmingen B KGV Submitted application in 1989Gundremmingen C KGV Submitted application in 1989lsar 1 Bayernwerk Submitted application in 1989Krummel KKK Submitted application in 1990Philippsburg I Badenwerk License not applied forWurgassen Preussen Elecktra License not applied forSwitzerlandLiebstradt KKL First MOX load in 1998Muhleberg BKW License not applied for[-lollandDodewaard GKN Has current MOX license

Cofrentes HE License not applied forSanta Maria de Garona Nuclenor License not applied for

2.3.4-4

It is clear from the previous summarythat in Finland, Sweden and Spain the utilities have not

taken any steps towards using MOX fuel; approachingthese utilities is thereforeof littlevalue. In

Germanyand Switzerlandsome utilitieshave applied to use MOX fuel, although none are likely

to be ableto load it early enough to be of use in the dispositionstudy, unless the licensing process

were to be accelerated. The Dodewaardreactorin Holland has already operatedwith MOX fuel,

but its smallsize bringsits suitabilityinto question.

Following this initial surveythe utilities at KKB, KKK, KGV, Bayernwek,.KKL and GKN havebeen contacted to establish their views on performing MOX lead test assembly irradiations. The

resultsof this are given below:

1. KKB and KKK (Brunsbutteland Krummel)

Contact:- Mr. Rigerof H.E.W.

•Mr. gieger stated that the Krummelreactorhas many differentfuel types within its core and

they are always interested in testing new fuel designs within the reactor. As such the

nsertion of GE MOX fuel would be an attractiveproposition to them.

However neither Krummelnor Brunsbuttel currentlyhave a MOX fuel operating license. A

public inquiryfor the Krummelwas planned for November 1992 but this was canceled and

to date there has been no furtherprograms. His view was that it is unlikelythat a MOX fuellicense will be obtainedbefore 1998.

He stated that H.E.W. were still interested and he has been requested to supply furtherinformation on the facilities available at the Brunsbuttel and Krummelsites.

2. KGV - (Gundremmingen B&C)

Contact:-Mr. Passig (RWE)

Mr. Passig stated that the license to load MOX fuel into Gundremmingenwas expected in

January 1994. However he did anticipate some problems in loading US DOE MOX fuel

assemblies into German reactors. The main problemfor RWE (and other German utilities)

was that they have to put priority on the utilization of plutonium arising from their

reprocessing contracts. Nevertheless he did state that he was interested in the concept andhis colleague Dr. Dibbert has been contacted with a view to provide furtherinformationon

the Gundremmingensites.

2.3.4-5

3. GKN - (Dodewaard)

Contact:-Mr. Van der Hulst

Mr. Van der Hulst stated that under normal circumstancesDodewaard would be ideally

suited to accept MOX fuel assembliesfor GE. However Dodewaard are experiencing

severe difficultieswith the Dutch licensingauthorities. Thishas resulted in a compromised..

'position being reached_underwhich a temporary-license.has been granted for continued

operationof Dodewaard. This license was restrictiveto the extent.that.GKV were ob!iged

to removeMOX fuel from the reactorandthey will only be allowed to load uraniumfuel inthe nearfuture.

4. BaYernwerk- (ISAR I)

Contact:- Mr. Huber

Mr. Huber stated that although a MOX operating license had been requested for ISAR 1

(BWR) Bayemwerk are not pressingfor its issue as they have decided to load MOX fuelonly into theirPWRreactorat ISAR2.

5. KKL- Liebstadt

Contact:-Mr. J. Afonso/Dr.Patak

Dr. Patak stated that the testing of leadMOX fuel assembliesfor GE would be an attractive

proposition to them. Mr. Afonso has been requested to provide additionaldetails of thefacilitiesavailableat the Liebstadtsite.

2.3.4-6

2.3.4.3 PIE Program

If a suitabledemonstrationirradiationin a commercialreactorcan be accomplished,it would need

to be followed by a PIE program. The aim of the program would be firstlyto confirmthat the

fuel had behavedsatisfactorily,and secondlyto studyindetailthose aspects influencedby isotopiccomposition. If rodscontainingburnablepoisonmaterialsare used then these wil!lalso need to be

includedin the PIE program. A typical programmightstudy perhapssix rods _nd consist of the

following:

Pool-Side Work

- visual examination

- ultrasonicleak testing

- measurementof assemblygrowth

- measurementof rodbowing- selection, removaland decontamination of chosen rods

- ECT measurementof clad oxide layer thickness

- transferof chosen rods to transport container

TransportfromReactor Site to Hot-Cell Facilities- hot-cell work

- fullvisual examinationof each rod

- measurementof rod growth

- gammascanni,_gof each rod

- profilometryof each rod

- puncture, gas analysis, internal pressure/volumemeasurementon each rod

- density measurementson selectedfuel samples

Sectioning at SelectedLocations;

- optical microscopy (showing grain structure, porosity distribution etc)

- alpha autoradiography(showing plutonium homogeneity)

- EPMA/SRF (givingelemental/isotopic distributions,fission gas retention etc)

- chemicalburnup analysis

Disposal of allWastes

2.3.4-7

2.3.4.4 Ramping Program

To assess the PCI resistance of a new fuel type, specialized ramp tests are required. Although

such tests may be judged unnecessary in the case of weapons MOX, if they were to be required, a

suitable test would comprise:

- transport of selected rods to hot cell facilities

- non-destructive and destructive examination of parent rods to include profdometry, gamma •

scanning, puncture testing and g_s analysis

- refabrication of each parent rod into typically three short rodlets_each rodlet to be pressurized

and sealed and neutron radiographed

- ramp testing of each rodlet, consisting of irradiationat a specified conditioning power for a

specified time, followed by ramping at a specified rate to a specified terminal power

...... _--_post-irradiationexamination of selected rodlets to include visual inspection, profilometry

neutron radiography, gamma scanning, internal volume/pressure, gas analysis, cerrnography

and density

- disposal of all wastes

2.3.4-8

2.4.1 FUEL FABRICATION FACILITYSYSTEMS REQUIREMENTS &ARCHITECTURE

Objectives:

A dynamic model of the plutonium inventory flow through the overall Pu dispositioninfrastructure is tobe developed with a focus on the Fuel Fabrication Facility (FFF), itsrelated accountability methods, and its external system interfaces.

The intent of planned model evaluations is to identifyaU system level customer and derivedrequirements which drive the. FFF functional-design and time related performance

" specifications. Then, dependent upon the Pu metal conversion and mixed oxide fabricationprocesses selected, alternate levels of process equipment replication to meet throughputrequirements, and all Special Nuclear Material waste streams with their relatedaccountability methods shall be evaluated.

The overall functional process timing model will be grown to cover the plutoniumdisposition from weapons to a DOE permanent disposal site for the irradiated reactor fuel.'Model details will be of a depth sufficient to address the FFF issues. The functional modelwill allow a consistent trace back to the driving systems level requirements, as well asdefining a consistent set of system interfaces and the allocation of functions to specificcomponents.

Status:

The functional model of the FFF has been defined and executed at a level of detailsufficient to identify and evaluate the pellet/rod storage requirements and the significant,fuel related scrap/waste streams. The RDD model is currently being checked for physicalconsistency and completeness. Once this task is complete, the design and manufacturingstaff will be engaged to assure realistic inputs to the simulation.

The FFF functional structure is based on past industry best practices, including maintainingphysical separation between the gadolinium and non-gadolinium fabrication processes.Hence, the FFF has a minimum of two independent fuel fabrication lines which, based ona GE9 bundle design, would be of equal capacity. The FFF design will be based on meetingthe ABWR reload core requirements for throughput; which can range from 324 to 972 fuelbundles per year, dependent on the reactor strategy selected (see section 1.0). To clarify theexternal interfaces and initiate the discussions with thedesign and.manufacturing staff wehave proposed an initial set of derived FFF requirements as follows:

• all materials required for fuel bundle fabrication shall be received in aform suitable for immediate use,

• all materials shall be made available as required to meet fabrication schedules,

• all process equipment shall be continuously monitored for acceptable performance

2.4.1-1

and, if necessary, adjusted on-line to maintain acceptable performance,

• process equipment lifetimes shall be greater than FFF usage requirements, or theirreplacement impact on fabrication schedules shall be minimized,

• all glovebox filter change operations shall be performed on-line,

• fuel accountability procedures shall be automated using current state-of-the-artequipment and processes,

• ..... fuel _accountability procedures •shall, not-havea.significant impact on fabricationschedules,

• fabrication process formation of fuel scrap material shall be minimized, and shall bebetter or equal to recent LWR fuel fabrication experience,

• fuel scrap wet recovery process shall y/eld a minimum of 99.5% of plutoniumback to the FFF,

• the storage area for completed fuel assemblies shall be sized for a full fresh core,

• FFF generated Waste shall be minimized

The current functional model of the FFF is presented in Figures la through lf, whichfollow the RDD decomposition of the system definition to the pellet fabrication process.The figures are the RDD-100 Behavior Diagrams (BDs), which are consistent with the database elements and the performance attributes. Table 1 provides explanatory notes for theBD' symbology as an aid in interpreting the functional model. The RDD model is beingchecked by simulating the production of a one fourth core reload, or 58 fuel bundles. Theemployed manufacturing strategy was to operate two independent pellet and rod fabricationlines, by campaigning the number of rods of each enrichment type, in ascending Puenrichment. Fuel bundles were _,ssembledas soon as the appropriate mix of rod types wereavailable. The GE9 fuel bundle is composed of 60 rods of the following types:

Rod Type Pu Content Gd Content Number of Rods

1 1.0% -- 4i i

3 1.6% -- 8

5 2.3% -- 8

6 2.8% -- 8

11 2.8% 1.0% 4

12 3.2% 1.0% 8

13 4.2% 1.0% 20

2.4.1-2

Sources of fuel-containing scrap were identified from previous LWR fabrication experienceand are indicated as the following mass fraction of fuel material throughput:

Pellet line operations: pellet pressing -> 0.008pellet grinding -> 0.024

Rod line operations: pellet loading -> 0.006

Glovebox coarse and HEPA filters, as well as vacuum bags were assumed to be consumedat fractional rates based on throughput of fuelpowder, and whenever a new enrichmentcampaign waslnitiated. A spent filterwasassumed to contain about 0.2Kg of fuel materialwhich was 99.9% extracted from the filter body before the filter was assigned to the lowlevel waste category (LLW).

The results of the simulation are shown in Figures 2 through 5. The top level results aresummarized by Figure 2. The curves labeled "generic pellets" and "generic rods" depict thetotal amounts in interim pellet and rod storage areas, summed over all fuel types. Thisparameter shall be used to size the FFF storage areas, and shall be the subject of futuretrade studies. The curves labeled "fuel bearing scrap" and "Total wet scrap" are the timedependent and integral of the fuel bearing scrap sent to the wet recovery process. Figures3 and 4 show the fabrication time lines for the different pellet and rod types, and their"generic" storage requirements.

The LLW curve values are presented in cubic feet, with compaction factors applied to thegenerated waste, which includes filters, bags, rags, Pu shipment cans, rejected fuel tubing,etc. The factors for compaction are currently arbitrary, and need to be revised. The timedependent, fractional life consumption of filters is depicted in Figure 5 for the mox fuelline, it is very similar for the gadolinium line. It is currently based on assuming that theglovebox operations generate 0.1% of the material throughput as spills or airborne dustwhich is collected by the filters and vacuum bags. Vacuum bags are assumed Io accumulateup to 0.25 Kg of fuel material during their useful life; realistic values will be determinedprior to any trade studies.

2.4.1-3

y !I

T. ""_' I Averticlelinerepresentsaprocess.i , ! _me_ downward,sequentially AIoopflowiseitheropenendedm toeachfuncl_n. orcontrolledbyanilera_logic,e _...do.,,

Funotlon In loop Oran'end_' exR, Ii I I ' ",no.. =or; i o_

_, Aparallelsetofprocesseshas tZ _, I • ! independenttimeflowwhichawaitsI ,.:-,I I--._.Z--Icompletionofeachprocessati I I, I itsconfluence.

i,Functions can outputitems which are either

real objects and/orinformation. Inputitems can be used totrigger functions.

jr 0._,)multi-exitfunctionisanOR output

branch,Theexittakenisdependentofthefunction

Table1,AnAidtoRDD=IO0BehaviorDiagramSymbology2.4.1-4

I decomposethisfunction I t

in Fig.lb

Fig.la. FunctionalModelfor:DisposeWeaponsPlutoniumviaReacto_r

newfuel reactorsite power "_make tritium

RequestZBUF 3.3

Output Tritium• Fuel Rx Electrical woduction

Pu02shpmt Fabrication Power (Option)

98

& ZBUFXfer HLWCask

J

decompose this function

in Fig. lc

Fig.lb-FunctionalModelfor:TransitionPu02toHighBurnupFuel

I0 _7

Io , ,_

decompose this tun_ioin Fig. ld

Fig.lc-FunctionalModelfor:FuelFabrication

I_iii::i::!i!!iiii!i!!!!...... :,,.. ,:,,!!!iiiii!i!!'.:,::!!iiI % i I n Itl m,I I_-e I

_iiii!i!ii i :ii_iiiilii!i!iiiiii iiiii::::!!i !ii!i_i_i/ m ox iI n _ I" : I .] il

813' _"

t ) _'---._ e.l .t.e... I

• ...... 18.1

.7 _ ..,,.: " )

_',i'_iii,,',!i__i',ii',ii',_ I_,__._I

\ \ __-..___ _,.,-.I.l.l.e... 11I.II _ 3 "1"1"e_', . _ 1e.I

\ i _,r,.,.,_'r°,,n,.r.w.,1 ----1,,,,,,oxt___.,r..r.__,.L____..___-I _ J I ,_,r _ _:._.'.'._I ,,,,,,,pies II '"_ '_N°x Il """" I

etorage

:1.1.1.¢1... fab.10

r_h,.,,,... ,,,.l

• ._._ .e... 21,1 _._ .1 .e... II1 .e

_,,d po,o, Request mox._..., _ .,.,.,,,,r, ,-...-.,.

Fig,lf- FunctionalModelfor:FabricateMoxPellets2.4.1-:10

I

ABWR Pu Phase Ic by Switick " 30_December 1993 at." 8.'22:46 am

Resource (LL'W pu/u vol)l20__

Item vs. Item View I Re'source (Total wet 's(:;ap)_ 10 -.-I.,500- 400 -- l 16 -4

14-.1450- , i 350- __12 -l4oo-] 300- _ _;lo 4 ....................._i-!ii:-!i!i!iii!iii_!i_i-!i!35o-} .-. " -"'___!_ _ 8 4 _i_':_::_;_'_:::.-=_i:_i

-_.oo--] _i50- -I _i!i:i-!i:i!i!i!i:iii!i_!i_!::_::-!i-!i!i!:i::_-. _'oo- 6 --t _!;!_!C!_!_!_!;!_!_!::!=_::!_ii!_::=--!i!::i_!::i::!_)_!ili!i!i!i!i_i_=,<.5 0 -_ ) " -- •===============================================================i _: =i=:::=::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::I : "=============================================:'::::::: =::::==:::==: 4 ==============================================================================

" =========================================================:: ===::=-:=:=: 3 ================================================================================:_15o_ ......................................................................................................................".................."........":......-,oo_ -,oo_. _i!iii!_i;i!!_iiii!_i;!i_i!_;ii!:_ii_!i_i{!_{!_ii!_i!_iiii!iii!__:_::i_i_i_i_i_i_iii_ii_!]_iii_ii_;ii_i_;iiiii_ii_!_i_i_i;4!ii_!_ii!_!i_i_i_;o _(, _i;ii!!:;i;;;i!;i;i;;;i_i;!;i_i_!i_i:i_i_i_i_i:i_i!_;_;_:_;_i!;_i;i:i:!;i;;_n_.oo ,_o _6oo2ooo2_oo

- 50- IRe_our_e(_02u_e_ca,,_)I50--__ 0 --" -1 160_ -0

0 _ _. & 4 _ 6 0 400 800 1200 1600 2000 2400| 4

iltem vs. Item viewI MOX rod type Resource. (MOX bearing scraPl, I / 120J ===========================================:=::==::===:::=....

_oo_. _,_ _!_:i!_i_!i_!!i!i_i_- _oo ...._:.:::i:i!:_:i!:.:.:_:_....

_i .:.:.v.:._.-:-=-:i:-=-:--:_-=::.:700 -4 22 _ =============.=====:=-:':=:=':=::'===.:::-1:======-========6oo-4 :_o

o_ " 18

:_ $.oo4 ] 14

• x - [ _10

,_ =0"00"4_ !IJi'l-I]_J JJ IIH]]]_JJ ll_JJ .I;_@i_!_!il_i_!i!!ii,; :i..,. _ _ ' ,_our¢_)O_i_i_i!!!_i_i_i_i!i_i!i_ii!i_i!i!!_i_iii_i_i!i_ii_i_i_i_i_i_i_i_i!iii!i_!_!i!i_i_i_ii!i_i_i_i!!!_iiii_i_i!!i_(FreshFuel _)undle)'rO 1600 20-00 24:00

"" 200--1, (5

1 O0 -4 2 60- !=_ . _

..=

0 '*10 11 12 13 1" 0 400 800 1200 1600 2000 2400 50 -J,

MOXG rod type time I40 ..4

-,i.

',30 ..4

-t.20 ..4

J10--4

-,,i.

1500 1700 1900 21 O0 2300 2500

[ time

Fig. 2- General Characteristics of RDD Simulation: Fabricate 1/4 Core Reload

2.4.1-12

2.4.1-13

2.4.2. Plutonium Feed Material Interface

As weapons are retiredfromthe existing nucleararsenal, the warheads are removed and shipped

to a disassemblylocation where the nuclearweapons components or "pits" are separatedfrom the

warheads and placed into storage. The plutonium in these weapons components is removed and

either refabricatedinto new warheads, stored as part of the strategic reserve_or identified as

excess and placed into storage for non-militaryuse. A plutonium-hydride/dehydrideprocess is

being developed and demonstrated by national labortorypersonnel for removing the plutonium

fromthe "pits"in retiredweapons. The four main advantagesassociated with this process are (1)

plutonium can be easily removedfrom the tight fitting"pit" configuration, (2) no additionalwaste

is generated by the processing, (3) the plutonium configurationcan be easily unclassified, and (4)

consolidation into metal ingots provides a smaller, more compact form for storage. In this

process, a "pit" is cut in half and each half is treated separately in a reaction vessel. A small

amount of hydrogen is introduced into the reactor chamber at an elevated temperature. The

hydrogen reacts with the plutonium, forming plutonium hydride which spalls from the "pit"

(volume change during the reaction) and falls into a collection vessel. In the process being

developedbyLawrence Livermore National Labortory(LLNL), hydrogen is added to the reactor

until all of the plutonium has been converted to hydrideand fallen into,the collection vessel. At

that point the hydrideis heated to drive off the hydrogen(dehydrided),the remainingplutonium is

meltedand cast into a metal button or ingot. In the Los Alamos National Labortory (LANL)

approach, uranium hydride is used to provide a small amount of hydrogen for reaction with the

plutonium to form the hydride. The hydride falls into a heated collection vessel where the

hydrogen is driven off and flows upwardto the top of the reactorhydridingmore plutonium. This

cyclic process continues until all of the plutonium is converted to hydride and spalls from the

"pit". The plutoniumin the collection vessel is meltedand cast to forma metal ingot. The reactor

vessels are relativelysmall for the hydride/dehydrideprocess and readilylend themselves to glove

box applications as well as automation.

The demonstration of the systems necessary for the disassembly of retired "pits" is being

completed in the Automated Retirement and Integrated Extraction System (ARIES) program

currentlyunderway at LANL. (Ref. 2.4.2-1) The conceptual design and prototype testing phases

of this activity have been completed and the long lead time equipment items have been ordered.

The need for the storage of the plutonium recovered from the retired"pits" has been identified

and LANL personnel are currently developing the requirementsfor ttfis facility. (Ref. 2.4.2-2)

This facility must be capable of handling the retiredweapons plutonium as well as plutonium

residues fromvarious weapons sites located thought out the country. As a result, four different

2.4.2-1

storage forms("pits", oxide, metal and stabilizedresidues) are being considered for this facility.

Ideally"pit" storage would be eliminatedfromthis facility because theirclassified naturerequires

additional security and control measures. The metal could be stored irrespective of impurity

content and the oxide would be accepted only after calciningso the loss on ignition at 1000 C

would be less than0.5 percent. As in the case of metal,no impuritylimits have been defined for

the Oxideform. The storagerequirementsfor thestabilizedresidueshavenotbeen finalized. -....

Both LLNL and LANL have indicated thereshould be no problemconvening either the hydrideor metal form of plutonium to an acceptable oxide, see Attachment 2.4.2-I for a draft

specification. Each laboratoryhas some historical experiencewith convening metal to oxide, but

additional testing will be requiredto evaluate the morphologyof the resultingpowder. Blending

tests are also suggested to evaluate the plutonium homogenity of the material resulting from a

physicalblendprocess. Although the presenceof galliumas an impuritydoes not appear to be a

problemfrom the neutronics standpoint,the effect of gallium on the physical behavior(thermal

conductivity,fissiongas release, irradiationdamage)of the fuel in a reactormust be evaluated if it

has not been removed during processing. If a problem is identified, technology must be

developed and demonstratedto remove the gallium from either the plutonium hydrideor metal.

Uncertainties concerning the powder morphology and the overall effect of the gallium in the

irradiatedfuel need to be evaluated andsolutions developedshoulda problembe uncovered.

The mixedoxide fabricationprocessbeing considered for the PlutoniumDisposition Study (PDS)

is based on proven glove box technology with automated operation and hands-onmaintenance.

In order to minimizethe worker radiationexposure during fabrication,the americium shall be

removed from the plutonium feed material. The hydride/dehydriderecovery process described

above contains a step for americiumremoval just prior to the casting of plutonium ingots.

Although this approach may be acceptable if the plutonium is fairly old because most of the

plutonium-241has decayed to americium-241,the lowest exposure plutonium dioxide feed would

result if the americiumis removedjust prior to oxide conversion. This scenario which results in

the lowest exposurefeed materialthat minimizesthe plutonium recyclingactivities suggests that

unless the plutonium is to be used as the oxide within a relatively short time, the plutonium

storage form that best fits with the PDS is unpurified plutonium metal. In this scenario, the

plutonium is removed from the "pits" by the hydride/dehydrideprocess, cast into ingots and

placed in storage. The plutoniumidentified for PDS activities is removed from storage, melted

and the americium removed. During these subsequent steps, any impurities could be removed

from the plutoniumby either electrorefiningor vacuumdistillation. The resultingplutoniumcould

2.4.2-2

be converted to oxide by an oxidation technique and shipped to the mixed oxide fabrication

facilityidentifiedfor the PDS mission.

References

2.4.2-1. W.Dworzak, al et., "ARIESConceptualDesign Report,"Los Alamos National .......

Laboratory,June 1, 1993, (NMT-DO:(U)93-041).

2.4.2-2. "PlutoniumStorage FacilityLead LaboratoryInterface Document," Los Alamos

National Laboratory, September, 1993.

2.4.2-3

Attachment 2.4.2-I

DRAFT SPECIFICATION

Pu02 Powder, Ceramic Grade- Dry Process

1.0 SCOPE

1.1 This specification establishes *,herequirements for ceramic grade plutonium dioxide

powder produced from plutonium obtained during retitrement of nuclear weapons.

2.0 APPLICABLE DOCUMEaNTS

2.1 The following publications form a part of this specification to the degree indicated

where applicable:

ASTM B214°1964 "Sieve Analysis of Granulated Metal Powders"

ASTM B329-1961 "Test for Apparent Density of Refractory Metals

and Compounds by the Scott Volumeter"

ASTM B330-1965 "Average Particle Size of Refractory Metals and

Compounds by the Fisher Sub-Sieve Sizer"

When the contents of this specification conflict with any document referenced herein,

the specification takes precedence.

3.0 GENERAL DESCRIFFION

3.1 Pow.der Description

The material to be furnished in accordance with the specification shall be plutonium

dioxide ready for fabrication of mixed (uranium-plutonium) oxide fuel pellets. The

supplier shall furnish all required reports and information defined in this

specification. The supplier shall perform and report all tests required by this

specification.

4.0 REQUIRF_aMENTS

4.1 Functional Criteria,

Purchaser will specify the nominal plutonium isotope compositions. The

requirements of this specification are for material in the form of homogeneous

powder. For this specification, homogeneity is defined as follows: A homogeneous

SpecificationRevision 1/11/94

2.4.2-4

DRAFT SPECIFICATION

Pu02 Powder, Ceramic Grade- Dry Process

lot shallconsist of all materialproducedfor a single shipment in a continuous processsequence; or it shall consist of a uniformblend of material.,producedin a batch-typeprocess; and it shall consist of an amountof material produced with constantprocessparameters. Successful processing of this material is dependentupon the consistencyof characteristics. This specification provides for a wide range of limits for certaincharacteristics. Once the supplier establishes a specific value tor each characteristicwithin the wide range, he shall maintain this value within the specified narrowrange.The limits of the narrowrange selected must fall within the wide range. The narrowlimits shall include both materialand measurementvariations.

4.2 productSpecificationAll plans and procedures for obtaining (1) homogeneity, (2) samples for chemical

analysis, (3) chemical composition, and (4) physical properties shall be submitted toPurchaserfor concurrence.

4.2.1 Powder Composition4.2.1.1 PlutoniumAnldysis

Plutonium minimum 85 w/o of plutonium dioxideIsotopic Content Pu238, 239, 240, 241,242 as specified

4.2.1.2 The total volatile content of the powder shall not exceed 0.05 w/o plutonium dioxide.The total volatile content shall be determined by weight loss on heating a one gram

sample at 1000 + 25 Celsius for four hours in air or by an equivalent method

approved by Purchaser and agreed to by the Supplier.

4.2.1.3 The plutonium shall be in the form of plutonium dioxide produced by oxidation ofplutonium metal at or below 500 Celsius.

4.2.2 Imt)urities

4.2.2.1 Total impurities, shown on Table 1, shall not exceed 2,500 ppm of plutonium.4.2.2.2 The americium content shall not exceed 300 ppm of plutonium.

4.2.2.3 The uranium content shall not exceed 2500 ppm of plutonium.4.2.2.4 The total measured impurities in Paragraph 4.2.2.1 shall be such that:

I;Ci(B=) is equal to or less than 4.0 ppm by weight of natural boron excludingAm-241 where:

SpecificationRevision 11/09/93

2.4.2-5

DRAFT SPECIFICATION

Pu02 Powder, Ceramic Grade- Dry Process

Table 1. TOTAL IMPURITIES

I ,, Illll Ill II

IMPURITY LIMIT PPM OF PuIllll I I I I IllUllll IllII I II III

Aluminum 400Boron 1Cadmium 1Calcium 250Carbon 500Chlorine 25Chromium 150Cobalt 75

Copper 400Fluorine 130Gallium 400Iron 400Le_d 400Maganese 200Magnesium 200Molybdenum 400Nickel 400

Nitrogen 100Silicon 200Silver 25Sodium 400Tin 400Titanium 200Thorium 10Vanadium 400Zirconium 400

ZDyspr0sium_ Gadolinium_ Europium_Samarium 2

SpecificationRevision 11109193

2.4.2-6

DRAFT SPECIFICATION

]Lh,02Powder, Ceramic Grade- Dry Process...w........

Ci ffi Weight fraction in units of parts per million parts of Pu of eachimpurityin Paragraph4.2.2.1, and

(13-)- The naturalboron equivalent of element i from the Table 2 list,which is based on the neutronabsorptioncross sections in BNL-325 S1, and "ResonanceIntegral Data," ANL Newsletter No. 1,assuming a Maxwellian spectrum in the resonance region for atypical thermal reactor. The atomic weights are on the physicalscale. When impuritiesare reportedas less than a stated thresholdof detection, the thresholdvalue shoalbe used for this calculation.

4.2.3 Physical Pro_vertie_4.2.3.1 All material must pass througha 325-mesh U.S. Standardsieve in accordance with

ASTM B214-1964, "Sieve Analysis of GranularMetal Powders".4.2.3.2 Bulk Density

The bulk density shall be no less than 1.0 gm/cc as determined by the ScottVolumeter per ASTM B329-1961, "Test for ApparentDensity of Refractory Metalsand Compounds by the Scott Volumeter". Once the Supplier has established a value

for bulk density, he shall maintain this value within +0.25 gm/cc for the totalamountof materialsupplied underthe purchasecontract.

4.2.3.3 Surfacf Are_

The surface area shall be determined by the Supplier by the B.E.T. method ofanalysis or an equivalent method approved by Purchaser. This requirement forsurface area shall not be the basis for the acceptanceor rejection of material, but isrequestedfor informationonly.

4.2.3.4 Particl_Siz_

The particle size shall be determined by ASTM B330-1965, "Average Particle Size of

Refractory Metals and Compounds by the Fisher Sub-Sieve Sizer" and/or equivalentmethod approved by the Purchaser. This requirement for panicle size shall not be the

basis for the acceptance or rejection of material, but is requested for informationonly.

SpecificationRevision 11/09/93

2.4.2-7

DRAFT SPECIFICATION

Pu02 Powder, Ceramic Grade- Dry Process

..................... , ._ - _

Table 2 BORON EQUIVALENTDATA

IIIIIII II IIIIIII I II

Boron Equivalent Boron EquivalentElement l_m B/ppm of Element Element ppm B/ppm of Element

I IIIIIIIII I I I I I IIIII I III IIII

'Aluminum 1.32x10"4 Magnesium 1.42x 10.4Boron ' i.00' Manganese ,,,i,,' 4101x10"3 .....Cadmium 7.79x 10-! Molybdenum ......1.32x 10-3 ....Calcium ..... 2.87x10 "4 Nickel l i20x10"3

Carbon 5.51X10"6 ...... Nitrogen ......... 2.76xi0 -3 ...... .......Chlorine 1.37x10 '-2 Samarium 0.524Chromium ' 9.35x10 "4 ' Silicon 8124x10"5 'Cobalt 1.25X10"2 ..... Silver ' 2,98X10'2........

'Copper 9.45x10 -4 ...... Sodium 3.37x10 "4Dysprosium 9.7x 10"2 ' Tin _, 2106x10"4

0.434 Tungsten' 7102x10-3Fluorine 3.48x10_ Vanadium 1.49x10 '3 ¢

, , ,,,

Gadolinium 4.191 Zinc 3.71x10 "46allium 4.000x10 "6 Zirconium 2.93X10-:5Iron ' 7.43x10"4 .......

,,, ,,,, , ,,,,

Lead 1.23x10°.......

4.3 QualityAssurance R_uirements4.3.1 Concurrence of Purchaser is required, where indicated in this specification, of the

Supplier's Quality Control Plans and Fabrication Procedures prior to the productionof any material. This concurrence may involve the witnessing of tests and test

equipment at the Supplier's plant. Any significant change in process operations bythe Supplier is to be made known to Purchaser to permit joint performance

evaluation. Such changes may be accidental or planned.

4.3.2 Check Analysis SampleThe Supplier shaU take a sample from each lot of material sufficiently large that the

Purchaserwill receive a 20 gram sample. This sample is to be divided into four parts

for (1) chemical analysis by Purchaser, (2) chemical analysis by Supplier, (3)

SpecificationRevision 11/09/93

2.4.2-8

DRAFT SPECIFICATION

Pu02 Powder, Ceramic Grade- Dry Process

chemical analysis by Referee if necessary, an, (4) archive sample held by Supplier,until shipper receiver differences, if any, are settled. Thesample for Purchasershallbe packaged,-marked "CheckSample"and sentto Purchaserwith the lot shipment.

4.3.3 Referee

In the event the Supplier and Purchaser fail to agree as to the material producedmeeting any attributeof this specification or other requirements of the order, theyshalljointly select a thirdpartyand the method and degree of retesting. The findingof the third party shall be final. The partywhose value is farthestfrom the referee'svalue shall bear the cost of the refereeinvestigation.

5.0 INSPECTION

5.1 The Suppliershall submit to the Purchaser(prior to shipment) a Certificateof Test intriplicate for each homogeneous lot of powder showing that the material conforms tothis specification. The certification shall include the purchase order number,

Purchaser specification designation, the results of the required tests, and a statementcertifying that the material is homogeneous. The test results shall be so numberedthat they can be identified with their related lot of material. The certifications shallinclude the results of the analyses per Paragraph 4.2, including the limit of detection

for each analysis. This section shall include calculated total boron equivalent per theequation in Paragraph4.2.2.4, if applicable.

5.2 Production material is subject to return at the Supplier's expense unless theCertification of Test accompanies the shipment.

6.0 PACKING, MARKING, AND CRITICALITY REQUIREMENTS

6.1 The material shall be packed in a manner which will prevent powder loss andcontamination spread during transit. Shipments must adhere to applicable UnitedState Department of Energy and/or United States Depa/tment of Transportationshipping regulationsand licenses.

6.2 The shipping containers will be decided upon jointly by the Purchaser and theSuppliersuch thatPurchaser'slicense is not violated. The Purchasermust be notifiedand approveall pending shipments.

SpecificationRevision 11/09/93

2.4.2-9

DRAFT SPECIFICATION

Pu02 Powder, Cermnle Grade- Dry Process

6.3 The external and internal surfaces of the outer container and the external surface of

• the inner'container-shall be.as _freefrom contamination as possible, and in no case

shall exceed United State Department of Transportationregulation for smearablecontamination.

6.4 The Supplier shall identify each containeras follows:Outer Container:

(a) Containernumbers, innerand outer, including permitnumbers(b) Purchaseordernumbers

(c) Type of materialand lot number(d) Metal weight of material(e) Shipmentclass (Fissile Class I, II, or III)(f) Radiationunits(g) Purchaser'saddress (to be specified on purchaseorder)(h) Two packing slips mustbe attachedwhich include all shippingpapers including

two courtesy copies of AEC transferdocuments, and "Certificate of Test"analysis

InnerContainer:

(a) Containernumber(b) Type of materialand lot number(c) Gross, tare, net and metal weights

Specifi_ltionRevision 11/09/93

2.4.2-10

2.4.3 MOX Fuel Fabrication Facility Layout

The configuration of manymodernMOX fuel fabricationbuildings being constructed in Europe

and Japanare multi-story structures. As a result, an alternativeMOX fuel fabrication facility

building layout was studied which assumed a multi-story building configuration as opposed to

the original low level two-story building layout presented in the May, 1993, ABWR Pu

Disposition Phase 1A Report. The multi-story buildingarrangementis shown in Figures 2.4.3-

1 through2.4.3-3. The buildingarrangementincludes three floors above grade and one floor

below grade with overall dimensions of 220 feet by 220 feet with a height of 105 feet. This

configuration was compared to the original Phase 1A layout shown in Figures 2.4.3-4 and

2.4.3-5 which consists of two floors above grade with dimensions of approximately 490 feet by

360 feet with a height of 50 feet. These layouts were compared to determine if there was any

significant economic advantageassociated with one layout over the other. Due to the pre-

conceptual nature of the building design, and the lack of definitive site specific geologic,

meteorological, and hydrological data, it was not possible at this time to develop detailed cost

estimatesof the two alternatives. However, some general observationsbased on experience in

the design of similar nuclearstructurescan be made.

The concept of minimizing building costs by minimizing the surface area of external walls due

to nuclear tornado missile and seismic design criteria for a building designed for nuclear

material confinement was believed to be valid for early "conservative" nuclear building

designs. However, development of actual missile effect experimental data and modern seismic

dynamic analyses methods have allowed external wall thicknesses to be optimized resulting in

external walls which are in the range of from one to two feet thick which is often equal or less

than the thickness of interior walls (due to shielding requirements.) For example, a recent

design of a Category 1 plutonium confinement process building to current DOE Order 6430.1A

tornado and seismic design standards resulted in external walls which were one foot thick.

This structure was the main processing building for the Special Isotope Separation (SIS)

Plutonium Laser Isotope Separation Project which was planned for construction at the DOE

Idaho National Engineering Laboratory. The final building configuration was approximately

370 feet long by 310 feet wide with a 50 feet overall height. The external reinforced concrete

walls of this building were nominally one foot thick, which was the same thickness and

construction as most internal walls. This relatively low profile two-story building constructed

at grade was considered optimum. The process material handling steps, the interface between

this building and the laser beam tunnel, the local site soil conditions, and the relatively low

2.4.3-1

seismic acc, lerations for both building and equipment designs due to low building elevations

were the criteria which resulted in the optimized design.

The ABWR MOX fuel fabrication facility process involves transfer of powders, pellets, rods

and bundles as well as analytical samples, scrap and waste material between process/support

system steps. In the low level building layout, most of the process and support systems are

located on the first floor and material transfers are made by cart or horizontal conveyor on the

same floor level. In the ease of the multi-story layout, vertical transfers between floors at

different elevations are required. With the exception of the possible use of gravity feed in the

powder blending and milling processes, these vertical transfers will generally complicate the

material handling operations. The transfer mechanisms and enclosures must provide proper

confinement of material during normal transfers or a dropping accident, assure criticality safety

and allow remote monitoring and communication between floors during transfer. In addition,

these transfer mechanisms must assure that material accountability is maintained and that the

transfer mechanism maintenance can be accomplished within the transfer space provided.

Interfaces with services, utilities and transportation vehicles provided between the ABWR MOX

facility and other facilities on the site are generally located at grade and therefore more

compatible with the low profile building. Other factors which could affect the optimum

building configuration include land availability, local site soil, hydrologic and seismic

conditions as well as site labor rates and craft productivity during construction. Therefore, it is

believed atthis time that the' building configuration can be optimized only when more detailed

design information is developed.

2.4.3-2

_. 220' -0"

I

! ISTAmS --I

zo CHEM/MET LAB

WSR & WASTE ii TREAT_NT

i I

BUNDLE RODSTORAGE POWDER RECEIVING STORAGE

i PU02 INTERIM STORAGE

.1_ PU02 J PU02

VAULT koui/_ 3!" VAULTI ELEVi_

I-LIMIT OFI CATEGORY 1

I STRUCTURE

SHOP INST. NUC SAFETY _JLAUNDRY OFF ICE

io OFF ICES LAB

F ILTERI POWDER MEN' S/WOMEN' S GENERAL STORAGESHOP WASTE RECEIVING CHANGE ROOMS AND SHIPPINGSTDRAGE TRUCK BAY

SECUR II'YOFFICES

..

BASEMENT FIRST FLOOR - "GRADE LEVEL

Figure 2.4.3-1 Mixed Oxide Fuel Fob BuildingBosement ond First Floor Levels

GD203 FAB

D

] I EXHAUSTHEPA FILTERi R_Su

HVAC & FILER ROOMGD203/_X SCAN (POTENTIALLY CONTAMINATED)

X-RAY/NDE[NSPT. _XF_ [_D D_--'_ U

HOT.INT. !--] B B _- I

TEST _ _ _ _ _ -L|MIT .CATEG_Yr-Io oF-! STRUC_RE

.... .... ....... ...............[ W ......MATERIAL TARGET OLD MAINT

CONTR_ / "G" AREA i

_CHANICAL SERVICEEOUIP_NT ROOM

GENERAL OFFICES,_DICAL & CAFETERIA r--1

I I

2ncl FLOOR 3rd FLOOR

Figure 2.4.3-2 Mixed Oxide Fuel Fob BuildingSecond ond Third Floor Levels

, 220'-0 _

GLOVEBOXHVAC & FILTER ROOM EXHAUST

(POTENTIALLY CONTAMINATED) HEPA FILTERROOM5

GD203/MOX SCAN HOT MFGX-RAY/NDE INSPT. MAINT. TEST

OI I

POWDER RECEIVING ROD I

PU02 INTERIM STORAGE STORAGE I GRADE

Puo2

|

SECTION A-A

Figure 2.4.3-3 Mixed Oxide Fuel Fob BuildingMulti-Story Bundlng Section

2.4.3-5

I1 150M ..__1

F ]HOT BUNDLE ASSEMBLY

M,tlNT. AND INSPECTION

CHiEM/M(T LAB _ & WASTETREATMENT

MFCTEST

m

L._ 150M ..._i

F

c-n r--1

(",4

r_ D D D D D D D D D D

n J"-'l n r-'-'_. HVAC & FILTER i_)OM _/_ J

_T,_Lv _T_AT_O_ ._-ICt.ovEooXEXHAUST _1HEPA FILTE _ ROOMS cjr_ jr

V

O3 MECHANICALSERVICEEQUIPMENT ROOM

f

0 10 20; ......... ; ...... _"1

SCALE:-ltTBO

Figure 2.4.3-5 Mixed Ox;de Fuel Fob Bund;ngLow Level Bu_d;ngSecond Floor

2.4.4 FIRST-OF-A-KIND TECHNOLOGIES

An evaluation of the MOX fabricationtechnology was conducted to identify those

aspects of the process or the associatedfacilities which might be construed as first-of-a-

kind technology. MOX fuel has been fabricatedin the U.S. in significant quantities in

the past and MOX plants are now under design or construction in several foreign

countries. An evaluation of the various MOX pellet fabrication processes, includingthe mechanical blend, sol-gel and coprecal processes were presented in the Phase 1A

report. The mechanical blendprocess was selected as the reference process. Many of

the foreign plants, those in United Kingdom and Japan for example, plan to use this

mechanicalblend or dry process for fabricating pellets from mixed oxide powder. Abrief re-evaluationof the alternateprocesses was conduct¢_to confirm the choice of the

mechanical blend process as the reference. This process was chosen as the reference

because of the process simplicity, extensive experience base and the potential to result

in the lowest levels of waste. Unless the initial feed material source is expanded to

include other plutonium materials than those obtained from retirement of nuclear

weapons, there appears to be no reason to change to an aqueous process to remove orreduce impurity levels.

The mechanical blend technology for MOX fuel is well established and no major

processing changes have occurred, significant advances in the areas of automation,

material handling, equipmentdesign, and real time instrumentationcan lower worker

exposure, improve quality and increase throughput. The application of the

advancements in each of these areas, with the possible exception of instrumentation

which is considered a developmentactivity, to MOX fabricationare considered first-of-a-kind type applications.

Since the normal processing operations in the MOX factory must be automated to

satisfy reduced worker exposure limits, automationtechnology must be integrated into

the material handling activities of the factory. The majority of this automation or

robotics equipment is currently available, but integration into the current process

equipmentarrangementshas not been accomplished. Many of the automationconcepts

being utilized by foreign fuel fabricators for processing of plutonium recovered from

spent fuel are directly applicable, but each installation is site specific and the details of

operation must be developed for each application. A good example of this type of

2.4.4-1

integration is the location of mechanical assists to remove or maintain an equipmentitem or modularcomponent.

Although the technology for conversion of the plutonium metal to oxide to provide the

feed material required for the MOX fabrication process in not considered part of the

activities to be evaluatedin this phase of the effort, the facility to produce this material

is considered a first-of-a-kind installation. The required conversion technology hasbeen developed for batch type operations during activities completed in the weapons

community. These processes need to be adaptedto produce the quality and quantity of

plutonium dioxide which meets the requirementsfor fuel fabricationin a MOX facility.A series of process demonstration studies may be required to optimize the conversion

processes to satisfy the requirementsof the plutoniumdisposition activities.

2.4.4-2

2.5 WASTESTREAMCHARACTERIZATIONANDMANAGEMENT

The operation of the plutonium disposition complexwill generate a volumeof

waste each year. The various waste types to be generated tnclude spent nuclear

fuel (or high-level radioactive waste), low-level radioactive waste, transuranic

waste, hazardouswaste, and solid and sanitary wastes. Waste activities are

regulated by a variety of government agencies as listed in Table 2.5.1.

Treatment, storage and disposal of generated wastes are controlled by a variety

of instruments such as licenses, permits, certifications, consent orders, or

other written approvals.

Table 2.5.1. WasteRegulatory Agency

_ Waste:Type { j Regulating AgencY

Spent Nuclear Department of Energy and/or NuclearFuel (High-Level Regulatory CommissionRadioactiveWaste) m,,

Low-Level Departmentof Energy, NuclearRadioactive Waste Regulatory Commissionor StaLe (if......... an agreementstate)

Transuran!c Waste Department of Energy

HazardousWaste Environmental Protection Agencyorcognizant State agency (if RCRA

..... programapproved)

Hazardousand Departmentof Energy and theRadioactive Mixed Environmental Protection AgencyorWaste cognizant State agency (if RCRA

program approved) i

Soltd and Local health agencystandardsSanttary Wastes

One of the primary design goals for the plutonium disposition complex is

minimizationof wastesgeneratedin the variousprocesses.These effortsare

definitelycost effectiveas the cost and treatment/disposaluncertaintiesof

handlingthiswasteincreaseinthefuture.Thissectiondiscussesthetypesand

characteristicsof wastesgeneratedineachactivityoftheplutoniumdisposition

2.5-I

complex. For eachwaste stream, the source of the generated waste is identified

and waste treatment and minimization activities prior to disposal are also

discussed. For clartty, this section begins with defining the various wasteforms dt scussed.

WasteOeflrlltton_;

Hiah-Level Radioactive Waste (HLW)-The highly radtoactive waste material that

results fromthe reprocessingof spentnuclearfuel.

Low-LevelRadioactiveWaste (LLW)- Radioactivewastethat is not high-level

wasteor containslessthan 100nCi/gramTRU concentration.

TransuranicWaste(TRU)- Radioactivewastewitha TRU (alpha-emittingTRUwith

half-livesqreaterthan20years)concentrationgreaterthan100nCi/gramofTRU

isotopesin the wastemass.

HazardousWaste- Wastesdesignatedas hazardousbyEPAregulations(40CFR261).

Hazardousand RadioactiveMixedWaste- Wastescontainingbothradioactiveand

hazardouscomponentsas definedby theAtomicEnergyAct and RCRA.

NonhazardousSolid Waste - Non-regulatedwaste (exceptfor local landfill

requirements)

Sanitar.yWaste - Waste water normallydisposedin a site drainfieldor a

municipalsewersystem.

Spent Nuclear Fuel - Fuel that has been withdrawnfrQm a nuclearreactor

followingirradiation,has undergoneat leastone year'sdecaysincebeingused

as a sourceof energyin a powerreactor,and hasnot beenchemicallyseparated

intoitsconstituentelementsby reprocessing.Spentfuel includesthe special

nuclearmaterial,byproductmaterial,sourcematerial,and other radioactive

materialsassociatedwith fuel assemblies.

2.5-2

2.5.1 MOXFuel Manufacturing Facility

Wastes from the Mixed-Oxide (MOX) Fuel Fabrication Factllty wtll be generated tn

the manufacture of the fuel assemblies and durtng ancillary acttvtt|es. The

prtmary waste types wtll be TRU and LLW, however, small amounts of hazardous,

soltd, and santtary wasteswt11 also be generated. AsdtscussedtnSectton2.4.1

the flOX factllty simulation studtes have not yet reached the potnt where new

waste stream Information ts available. Therefore the results provtded for the

flOX factllty tn thesummary tables were taken from the Phase IA studies.

2.5.2 ABWR

Wastes from the Advanced Boiling Water Reactor (ABWR)will be generated tn the

operation and maintenance of the reactor and during ancillary activities. The

prtmary waste types will be spent fuel and LLW, however, small amounts of

hazardous, solid, and sanitary wastes will also be generated. The following

discussion provides specific information on the ABWRwaste streams. For

simplicity, the discussion wil1 be directed towards waste produced by one ABWR

unit. The total waste volume for the two unit plutonium destruction complex willbe tabulated at the end of the discussion.

2.5.2.1 High-Level Radioactive Waste

The operation of the ABWRwill produce spent nuclear fuel (SNF). Although not

generally considered the same, this report will use the terms high-level

radioactive waste (HLW) and spent nuclear fuel interchangeably. Using strict

definitions, the ABWRproduces only SNF; no HLWis produced as SNFwi11 not be

chemically processed prior to disposal. The disposition of this waste is the

responsibility of the DOE. The amount of SNF waste generated annually is

projected to be 162 bundles (29 MTHM)per reactor, based on a 75% capacity

factor and a 523 day cycle. The ABWRMOXfuel would be irradiated to a burnup

of 38,000 MWd/t which is typical of the SNF generated by light water reactorslicensed in the United States.

2.5-3

Treatment;

Onstte- Spent fuel wtll be stored in the reactor spent fuel pool for a minimum

cool-downperiod prior to further storage on site or shipmentfor offstte storageor disposal. There wtll be no other treatment of this waste at the reactor site,

except for the removal of target rods in the tritium production mode.

'Offstte - No offstte treatment of the wastets_projected at thistime except for

the placementof spent fuel bundles in a repository-required container prior toplacement in the repository (refer to section 4.4.4 for additional information

on repository requirements).

Disposal

.... : Disposal of SNFis the responsibility of the DOEper the Nuclear WastePolicy Act

of 1982, as amended. Current plans are to ship SNFto a national repository for

long term deep geologic storage/disposal.

2.5.2.2 Low-LevelRadioactive Waste

The operation and maintenanceof the ABWRwill produce a quantity of LLWeach

year. The total amountof LLWgenerated annually is projected to be 165 m3. The

disposition of this waste is the responsibility of the DOE. LLWwill either be

treated within the complex and disposed on the same DOE site or will be

transported to another DOEsite for disposal. Dependingupon site specific

acceptance requirements, the waste maybe grouted, compacted,vitrified, and/or

otherwise treated prior to final disposal. LLWwill exist in three physicalforms (solid, liquid, gas) as discussed below.

SolidRadioactiveWaste

Solid radioactivewastesare generatedprimarilyin thevariousreactorfluid

cleanupsystems. Table2.5.2.2,-Iliststhe major contributorsto the waste

2.5-4

volume from "wet" waste processes. After processing, the resulting waste forms

are a solld andwill meet dtsposal crtterta with respect to free l|qutds.

Table 2.5.2.2-1. Wet WasteGeneration - Pretreatment

. m . , _ . _m .... m m r_m]mm m _ _ m

i_::_iiiii_::.,_Vol_;Generated. _iii__SPecitic i-,__.iii_

I

CUWF/D Sludge.... 4.7 7.3 E7

FPCF/C Sludge 1.8 ....... 1.94 E6

CondensateFilter S!udge 4.6 2.40 E5iii i i,rl i

Leg Ftlter Sludge 0.2 1.5 E6i i i J ,1

CondensateDemtnerallzer 18.0 6.7 E4Resin

i ii i

LCNDemineralizer resin 5.0 1.18 E5ii i i

HCWDemtneraltzer resin 2.7 8.4 EO

Concentrated Liquid Waste 27.4 4.67 E3

LTotal 64.4 ......,I

Table 2.5.2.2-2 provides a projection of the volumes(pre-treated) andactivitiesof dry LLg generated by the ABWR.Table 2.5.2.2-3 is a summaryof the treated

waste volumesandactivities that are projected to be shipped for disposal aftertreatment.

Liouid Radioactive Waste

The radioactive waste treatment systemswill generate about 27.4 m3 per year of

concentrated liquid waste. This waste will be solidified in preparation foroffsite shipment and disposal. Therefore, no liquid radioactive waste would

require disposal.

GaseousRadl,oactive Wal_te_s

2.5-5

Theprimary sourceof radioactive gasests generation during the flsston process.

A small amountof noble gases is released tnto the coolant andthen collected by

the offgas treatment system. This systemis basically a holdup that delays the

release of nobel gases a11owtng for a period of radioactive decay prior to

atmospheric release. The flow capacity of the system is 40 m3 per hour. Holdup

times are a minimumof 30 days for Xenonand 40 hours for Krypton. Due to the

relatively short half-life of mostnoble gases, no gaseouswaste is collected fordisposal.

Table 2.5.2.2-2. "Dry" Soltd Haste - Pretreatment

Dry Waste Source VolumeGenerated Totalm3/Year Curiesi i i

i I iiiimmli

Combustible 225 1.6Waste

Compactible 38 0.3Wastei1,11

Other Waste 100 7.0

Total 363 8.9

Table 2.5.2o2-3. Annual ShippedWasteii iiii ' i ii i

Haste Type ShippedWaste Total CuriesVolumem3/Year

i I

Concentrated Waste 4.4 1.3ii ,

Combustible Waste 5.6 1.6iii i

CompressibleWaste ]5 0.3

Resins and Sludges 40 670Other Waste 100 7

i

Total 165 680.2

The reactorfacilityventilationsystemalsocollectscontaminatedparticulates

inthe air. The systemcollectsairbornecontaminantsin variousareasat a rate

dependentuponthe potentialfor and severityof contamination.The collected

2.5-6

airis thenpassedthrougha highefficiencyparticulateair (HEPA)filterwhere

it is cleanedof 99.95%of particulates.Aftera limitedlife,the contaminated

HEPAsare replacedand disposedas solidradioactivewaste.

2.5.2.3 HazardousWaste

' The operation and maintenance of the ABWRwtll generate a small volumeof waste

considered to be hazardous 'by EPA regulations (40 CFR 261). Examples of

hazardouswastes are paints, solvents, lubricants, film developmentsolutions,

laboratory chemicals, andother chemicals that cannot be disposed as solid non-

hazardouswaste. This waste is expected to be transported offstte for treatment

anddisposal by a contractor. This waste is expected to total about 290 m3/year.

2.5.2.4 Non-HazardousSolid Waste

The presenceof the ABWRwork force will result in the generation of a volumeof

non-hazardouswaste. This waste consists of garbage, office trash, packing

material, and other refuse. A disposal contractor will haul the 3,100 m3 per

year of waste to a local landfill for disposal.

2.5.2.5 Sanitary Waste

Sanitary waste is that liquid stream normally disposed in a site drainfteld or

a municipal sewer system. The presence of the ABWRwork force will result in

about 1.6 E6 gallons per year that need disposal.

2.5.3 Tritium Production

Wastesfrom the manufactureof targetrods and from the extractionof tritium

will begeneratedin relativelysmallquantities.No radioactivewasteswillbe

generatedin the manufactureof targetrods. The primarywastetypegenerated

duringextractionis LLW,however,verysmallincrementalamountsof hazardous,

solid,and sanitarywasteswill alsobe generatedduringbothactivities.The

followingdiscussionprovidesspecificinformationon thesewastestreams.

2.5-7

2.5.3.1 Low-Level Radioactive Waste

Tritiumproductionactivitieswouldadd a verysmallincrementalamountof low-

levelradioactivewasteto the complextotal. Thiswastewouldbe generatedat

the reactorsite primarilyduringthe removalof irradiatedtargetrods from

spent fuel bundles. This waste has basically the same_ _ :_,_':_::cs as other

LLWgenerated in the reactor refueling area and the ;_ _ ,._ _rage_pool

area. This waste would consistprimarilyof used protec _ .... r,,,ig andother

disposable items that have comein contact with pool water.

A smallamountof LLW would be generatedfrom the spent targetrods after

tritiumextraction.The spenttargetrodsare expectedto containabout50 Ci

each of residualtritiumand shouldalsocontaina smallquantityof activation

' " productswithintherodmetals. Mostof theactivationproductshaveshorthalf-

lives,but traceamountsof Co-60willcausean externalradiationhazardfor a

relativelyIongperiodof time. Spenttargetrodsshouldbe classifiedas LLW;

probablyas ClassB (10CFR 61.55). However,they couldalsobe consideredas

spent nuclearfuel by one NRC definition(10 CFR 72.3)whicl;would require

geologicrepositorydisposal.At thistime,the mostreasonableandeconomical

disposaloptionwouldbe as I'"

_Thesetwo activitiesare pr_v _edto generateapproximatelyI0 m3/yearof

additionalLLWvolume.Thisvolumeof LLWisapplicableonlyiftheABWR isused

in the tritiumproductionmode and wouldnot be generatedif tritiumis not

produced. The low-levelradioactivewastewill be disposedat a LLW disposal

site. This sitemay be at the samelocationas the complexor the wastecould

be transportedto the nearestDOE LLW disposalsite. Dependingupon site

specific acceptancerequirements,thewastemaybegrouted,compacted,vitrified,

and/orotherwisetreatedpriorto finaldisposal.

2.5.3.2 HazardousWaste

Tritiumproductionactivitieswouldadd an insignificantincrementalamountof

hazardouswasteeachyear.

2.5-8

2.5.3.3 Non-HazardousSolid Waste

Tritium production activities would add an tnsigntficanL incremental amountof

non-hazardouswaste each year.

2.5.3.4 Sanitary Waste

°

Trttium prOduction activities would addan insignificant incremental amountof

sanitary waste each year.

2.5.4 SuM.ary

This section described the characteristics, amounts, treatment, and final

" disposition of the wastes generated in the operation of the plutonium disposition

complex.This information is tabulated in Table 2.5.4-1 and is valid for a HOX

fuel manufacturing facility, two ABWRs,andancillary facilities in the complex.

The information provided indicates that the wastes generated are no greater in

volumethan for other similar projects.

Table 2.5.4-1. ComplexWasteStream Summarym,

i Waste........ .... Principle Quantity Treatment DispositionStream Source Per Year

- iI

MOXFacility ......

Solid Compaction DOE LLWRadioactive 60m3 disposalsiteWaste

i

Low-Level Liquid SolidificationN/AWaste Radioactive - 0

Wastei.,

Gaseous N/A HEPA Atmosphericeffl uent f i l trat i on re l ease

i i i iiii,

Trans- Fuel Included in WIPP Transport touranic Manufacturing LLWtotal certification WIPPWaste Process

2.5-9

;i_ ii!iPrinctple ii i _ Quanttty Treatment _D|sposttton;ii_iiiiiiiiiiStream_iiii_i__:ii!iiii_!Source /i;i !ii__I__!;Per Year i_;_!i! _i ;_

Hazardous Fuel No onsite ContractorWaste Manufacturing lOOm3 treatment disposal per

- Process . RCRA

Solid Non-hazardous 30m3 No onsite Disposal inWaste waste sources treatment local landfill

i i i

Sanitary Sanitary No onstte Disposal perWaste sources 2.1 E6 9al treatment local codes

I I

High-Level Spent Fuel 324 Decay >1 year HLWrepositoryWaste Bundles

58 MTHM

Solid 330 m3 Evaporation DOELLW• Radioactive 1360 Ci Compaction disposal site

Waste

Low-Level Liquid None Recycled toWaste Radioactive discharged Demineral izer condensate

Waste storacje

Gaseous N/A Offgas holdup Atmosphericeffluent HEPA rel ease

filtration

Hazardous Maintenance No onsite ContractorWaste activities 580 m3 treatment disposal per

RCRA

Solid Non-hazardous No onsite Disposal inWaste Waste-Sources 6,200 m3 treatment local landfill

Sanitary Sanitary 3.2 E6 Gal. No onsite Disposal perWaste Sources treatment local codes

m

TrtttumExtraction ..

Low-Level Spent target 10 m3 Compaction DOE LLWWaste rods disposal site

Solid Non-hazardous ~0 No onsite Disposal inWaste waste sources treatment local landfill

, i

Sanitary Sanitary ~0 No onsite Disposal perWaste sources treatment loc._lcodes

i "'"

2.5-]0

3.0 TRITIUM PRODUCTION

In this section, the option to convert the reactor system to produce tritium has been evaluated.

Work during Phase 1C has led to a MOX fueled core design that meets the tritium contract

quantity requirements. This Phase iC core design has a core average Pu enrichment of 5.9%

with an average burnup of 28,000 MWD/MT. All nuclear and thermo-mechanical design

criteriafor normal operation have been met. There are four tritium target rods per assembly

(similar to-1A). The currently designed core assumes that target rods are irradiated for one

cycle and then processed.

The ABWR can produce the contract quantity of tritium using partial core reloads and longer

target exposure time. However, the target rod performance data is currently limited to about one

year exposure. While we believe the target rod is capable of much higher exposures in the

ABWR, the reference design is based on discharging all target rods after one cycle to provide a

conservative basis for initial operation.

Additional evaluations of the worker and public health impacts of tritium production in the

ABWR confirmed that these impacts are minimal and that no plant modifications would be

required. The average dose to a plant worker is 1/100 of the DOE limit and the maximally

exposed offsite person is 1/1000 of the EPA limit.

The key difference between the two designs is the peak clad temperature during the design basis

loss of coolant accident (LOCA). This was the limiting factor in the NPR reference design

whereas the ABWR target rod isn't subject to any significant temperature transients.

The equilibrium core design analysis for tritium production is given in Section 3.1, followed by

an examination of the target rod design and performance in Section 3.2 The support facility

requirements are given in Sections 3.3. The impact of tritium production on ABWR plant

operations is examined in Section 3.4.

3.1 MOX CORE DESIGN FOR TRITIUM PRODUCTION

Phase 1A discussed a urania ABWR fuel bundle and core design for tritium production. This

section reports a MOX tritium production core capable of producing 43 million curies of tritium

per year. The tritium is produced inside specially designed rods containing a lithium-6 target.

3.1-1

Four of these target rods replace fuel rods in each fuel bundle in the core.

The basic concept calls for an in-reactor target residence time of about one year, after which

time the target rods are removed from the fuel bundles and sent to a tritium extraction and

processing facility. The important features of the core design for the reference tritium produc-

tion case are summarizedin the following sections.

..... : The reference tritium production core and fuel design can produce 43 million curies of tritium

per year in a single ABWR power plant. A summary of the important fuel cycle parameters and

tritium production rates is shown in Table 3.1-1.

Table 3.1.1. Parameter Summary, Tritium Case

Numberof Reactors 1Cycle Length, EFPD 320.9

Discharge Exposure, MWd/MT 28164Reload Batch Size 280

Average 239pu Enrichment 5.92%Plant Capacity Factor 87%Tritium Output, curies/yr 43xl 06

3.1.1 Reference Bundle Design

The bundle design for the tritium production option contains four 6Li target rods and 56 PuO2

power producing rods. This bundle design resembles the reactivity characteristics normally

associated with enriched uranium fuel. However the neutron absorption of the target rods nearly

doubles the enrichment required to obtain the fuel discharge exposure.

The bundle axial and radi',d enrichment distribution is given in Figure 3.1-1 along with the

gadolinia distribution. Values of enrichment for plutonium are read in hundreds of a percent.

For instance, the "310" refers to 3.10 w/o PU _'39. For the tritium production case, natural uranium

blankets are not used. This allows a maximum bundle average enrichment to overcome the 6Li

target rod reactivity penalty and also flattens the axial power shape so as to reduce the exposure

peaking in fl_etarget rods. There are a total of five pellet types used in this design, one of which

is a gadolinia bearing pellet.

3.1-2

Figure 3.1-1 Bundle Design for Tritium Productionip..-

iDee e o® @_ 0 0 _(b (D® @@@@@I_ (b

f3_

\2-"

©®_ @@@(- ""II 11

_ j

(9@@ @I

61..i

Urania • 071 071 071 Targel 071

Plutonia > 310 350 400 900

Gadolinia > 05o

Figure 3.1-2 Beginning of Life Assembly Power Distributionfor Tritium Production Assembly

.249 1.212 1.223 0.099i

I E938 0.96; 0.963 1.208

I ._82 1 3.964 1.193,,=. II.=_lm_=

b =23.934 1.227

I 2 3.934 1.227

I . 1 3.963 1.193.,,

' ,3 3.!_58 1.205

0.09911o20811,,193 93 1 ._05 0.099

..... [ [

3.1-4

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The infinite lattice radial power peaking is essential in determining the peak power producing

rod. The distribution of relative power peaking (normalized to unity across the lattice ) is shown

at beginning of life for the 40 percent void case in Figure 3.1-2. The exposure-dependent k.,

are given in Figure 3.1-3 for the uncontrolled lattice. This figure shows the k.. for three void

histories.

3.1.2 Equilibrium Core Design

The equilibrium core design philosophy for the tritium production option was to simulate the

reactivity distribution of a annual refueling equilibrium UO2 core in order to provide simple

operation. A single fuel nuclear design was utilized in an equilibrium batch of 280 bundles.

The detailed core design layout is presented in Figure 3.1-4. The numbers shown in the

beginning-of-equilibrium-cycle core map represent the relative number of cycles since fuel

loading. For instance, the number "1" refers to fresh fuel (loaded this cycle) and the number "4"

refers to bundles which are about to start their fourth cycle.

A single nuclear design of fuel is loaded into the equilibrium cycle. The important fuel bundle

parameters were summarized previously. A control cell core loading strategy that contains 37

control ceils was utilized. Due to the improved hot to cold reactivity swing characteristics of the

ABWR core, it was possible to design the fuel with a sufficient cold shutdown margin and still

maintain sufficient hot excess reactivity. The necessary high fissile content dictated a power

derate for the latter part of the cycle. Figure 3.1-5 shows the power profile through the cycle.

The important parameters of the equilibrium cycle design are summarized in Table 3.1-2.

Examination of the results reveals that all thermal and reactivity requirements are satisfied.

3.1.3 Core Thermal Margins

The critical power ratio and MAPLHGR thermal margin performance.are plotted as a function

of cycle exposure in Figure 3.1-6 and 3.1-7. Operation within the MAPLHGR limit assures the

mechanical integrity of the fuel rods is maintained by limiting their power output in an

appropriate manner throughout their lifetime. The MAPLHGR limits imposed on this cycle with

a relatively low discharge exposure are the same as the fuel licensed for up to 38 GWD/MT.

These results demonstrate ample mmgin to core thermal limits.

3.1-6

Table 3.1-2. Equilibrium Cycle Key Parameter SummaryTritium Case

Cycle Length, EFPD .... 320.9Cycle Energy, GWd 1260_

Cycle Exposure, MWd/MT 7818Core Mass, MT 146.4Reload Enrichment, w/o Pu-239 5.92Reload Batch Size 280Maximum MAPRAT 1.00Maximum CPRRAT 0.82

Maximum LHGR, KW/ft (LHGR limit = 14.4) 13.3MCPR (OLMCRP=l.25) 1.53Minimum Cold Shutdown Margin 1.60Hot Excess Reactivity at BOC 1.00

3.1.4 Reactivity Limit Summary

The reactivity performance of the tritium production option design is summarized in Figure 3.1-

8 and Figure 3.1-9 Due to the improved hot to cold reactivity swing of the ABWR N-lattice,

there is sufficient cold shutdown margin; therefore, there is little or no impact of the core design

from cold shutdown margin considerations.

Figure 3.1-4 Equilibrium Cycle Loading Pattern

{"-'-'2 - Second Cycle

CycleLmded 3 - ThirdCycle4 = Fomh Cycle

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3.1-8

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Figure 3.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure

EquilibriumCycleMaximumCPRRATvs. Exposure

1.1 I _ .............. ; - _--- ! .........

- _ i = =-1.05 - ............................_......................................................;........................ _..........................i..........................T...................... i.........................: = .......

i i i ii i iiiii -- iiiii ] ..../ i sm

/ : | iI i : ! i

0.95 1-........................................................_.........................._...............................'..........................._.............................._...........................*.............................i j i i i

/ t . ! i ! * !/

0.9 4-............................._................................_..............................._................................_.............................i..............................._............................;................................./ ; i I , i i/ i l _, I ; i i

t" I t _ ! ! t t,. 0.85 ..........................._................................'............................i..........................................................i.............................'..........................'.........................E: / _ _ ! -- I I i !

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o.751..........................!....................................................................... ............................0.7 .L ........................._.........................._....................._.....___ ......

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i i ; i i i___u._ ..........................*.............................-:............................_...............................i...........................+............................._.............................._.............................

o.6.. 1 _ ,. i ..........1, .... ._ i . _--_ ......0 1 2 3 4 5 6 7 8

Cycle Exposure {GWDIst)

3.1-11

Figure 3.1-7 Equilibrium Cycle Maximum CPRRAT vs Exposure

EquilibriumCycle Maximum CPRRATvs. Exposure

i i : i

I * i i i i ;

' [ i' | ill i ' ii i i t _ ii _ ii ....i ! : : i " i I

I ; i i i i i ;

v -- ..t_J. _(_J_ ........................... 4 ................................ .+ ............................ i................................ i............................. i............................... . .............................. . ...............................

0.9 ................................_................................._.................................._.................................i................................i................................4...............................i.....................................; ; } t i 1 i

o.85 I .............................._.............................._'............................._.................................'..............................}..............................._'..............................."...................................0.8 i , : - III _ i i 1 i

t ........................... ,_...............................t ..............................i........................ • ..........................i................................4-.................................+..................................

0.75

0.7 ...........................i................................_,..................................i...........

°°'t...............................i..............................................................i.................................i..............................i...............................,i................................i................................0.6

0 1 2 3 4 5 6 7 8

Cycle Exposure (GWD/stl

3.1-11

Figure 3.1-8 Equilibrium Cycle Hot Excess Reactivity

Equilibrium Cycle Hot Excess Reactivity

1.25 "

t,,-,,,. i

m 1 _'"

_._(D

"o

0.75 ........................................................................................................!............................................................................................

{ i ;

t,)n- 0.5

u>¢

L__" 0.25

0 ..............

0 1 2 3 4 5 6 7 8

Cycle Exposure (GWD/st)

3.1-12

Figure 3.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin

EquilibriumCycle Minimum Cold Shutdown Margin

3 i i _

.c== 2m:zC . : : i I

1.5 ....................................................................................................i.........................._.............................................................................................................................................0 : i ;

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o(J

0.5 .................................................................................................................................._,............................._......................................................................................................

0 ,_ l t

0 1 2 3 4 5 6 7 8

Cycle Exposure{GWD/st)

3.1-13

3.1.5 Core Performance Description

The core performance characteristics as a function of exposure through the cycle are given in

Figure 3.1-10 through Figure 3.1-13. The core maps in these figures show the control blade

patterns in the core expressed in terms of notches (which are three-inch sections of blade)

withdrawn from the top of the core. Those cells which have no numbers represent cells in which

there are no blades inserted. The thermal limits and reactivity mat'gins associated with the given

exposure are noted in the summary included with each figure. As seen from these figures, all

thermal and reactivity margins are met. The resulting core average power and exposure profile

are also given. Since the reactor core design itself provides sufficient margins, it is not

necessary to axial grade the fuel assembly to flatten or accommodate the shifts in power.

However it may be advantageous to flatten the axial power shape in order to reduce the exposure

peaking of the target rods. Table 3.1-3 shows target rod design parameters and performanceresults.

3.1-14

Table 3.1-3. Tritium Core Design Options

Target rod OD (in) 0.483Target rod clad thickness (in) 0.030

Target pellet OD 0.390Target pellet ID 0.2406Li enrichment 50%

Tritium Output 43.2 MCi/yr

6Li n-oc rates (per second per cm rod length)

BOC Average 2.7xl 0t3BOC Peak 5.4x 1013

MOC Average 2.7x10 _3MOC Peak 5.1x10 _3

EOC Average 2.7x 10_3EOC Peak 5.1x1013

3H Inventory (Ci per rod)BOC Average 0BOC Peak 0

MOC Average 6547MOC Peak 7535

EOC Average 12373EOC Peak 14070

End of Life Average 12373

Target Pellet Exposure (GVR)End of Life Average 147End of Life Peak 166

3.1-15

Figure 3.1-9 Equilibrium Cycle Minimum Cold Shutdown Margin

I =1

_ 36i 28 34 28 36 CycleExposure,MWd/st 0

36 24 32i 20 32 ..... 24 36 Cycle Energy, MWD 0Numberof Full PowerDays 0

- - .... CoreAverageVoidFraction .4572

28 i32 20 !30 ......20 _32= 28 "--1 Core Flow_ Mlb/hr I. 151ChannelPealdng 1.2654Maximum34 20 30 14 30 20 134 D CoreAxialPowerPeak ].30]

._d MaximumRAPLHGR .95628 32 20 30 20 32 28 M&gimum CPRRAT .819

• Hot ExcessReactivity, % 1.0536! 24 32J 20 32 24 36 Coid Shutdown Margin 2,84 ,

36 28; 34 28' 36

N N - Numberof 3 inch increments thatthe control bladeis withdrawnfrom fully inserted

26 28

2422 2220 2018 18

-_ 12 -_,.12 ................................................................

10 i _ 108 _ 8

6 i _ 6 i4 .4.,

4 _ i2 i _ 2 :

0 _ 0

0 G4 Q8 12 1.6 2 0 4 8 12 16 23 24 28

CoreAxialAverageRelative CoreAxialAverageExposure,Polar G1/_/st

3.1-16

Figure 3.1-11 Equilibrium Cycle Core Data Summary at 4000 MWD/st

" Cycle Exposure, MWd/st 4000

36 28 36 ,28 36 CycleEnergy,MWD 644660 ,NumberofFullPowerDays 164

, CoreAverage Void Fraction .47436 22 34 24 34 22 36! Core Flow, Mlb/hr 1.151 ..

,,MaximumChannel Peaking ,1129828 34 r26 34 26 34' 28 Core Axial Power Peak 1.40 -

"3 Maximum'RAPLHGR .982

36 24 34 26 34 24 36 . I MaximumCPRKAT .797_..,'i Hot Excess Reactivity, % 0.63 ....

28 34 26 34 26 34 28 Cold ShutdownMargin 1.6 _

36 22 34 24 34 22 36i

36 28 34 28 36

N - Numberof3 inchincrementsthatthecontrolbladeiswithdrawnfromfull)'inserted

2626 i _ i " ' -

24 .......................i................

18 18

10 _ 16

14 7=-_ 12 _..................+.................. -_ 12

1o8 8

6 ....................... 64 4

2 20 0 ! i ! -! ! i,,,

0 Q4 08 12. 1.6 2 0 4 8 12 16 29 24 28

CoreAxialAverageRelalive CoreAxialAverageExposure,Po_er G'V_/st

3.1-17

Figure 3.1-12 Equilibrium Cycle Core Data Summary at 6000 MWD/st

34 36 34 - MWd/st 6000..... Cycle Energy, MWD 966990

- Number of FuUPower Days 24636 34 36: 34 36 "Core Average Void Fraction .480Core Flow, Mlb/hr ' 1.151 .......

34 34 36 32 361 34 34 Maximum Channel Peaking 1.271Core Axial Power Peak 1.40

36 36 j32 32 36 36 _Maximum RAPLHGR .938 -,,

Maximum CPRRAT .777

34 34 36 34 36 34 34 -Hot ExcessReactivity, % 0.10

Cold ShutdownMargin ].6 ....36 34 36 34 136

34 34 34

N - Numberof 3 inch incrementsthatthe controlbladeis withdraw_fromfullyinserted

26 28

24 ................................................................,..................... i

i i _ i : i! ! _ ! ; i

16 _ 16

z_12 ....................i......................_....................................4................... ,.= 12

J

Idlam ! i =

8gJ ................... _................... _ ................... i................ ._...................

.................. i ..................... ! .................. i................. _ ..................

" ............... i..............i...............2 0 4 i 1 i 1 1

0 i ..... 4 l l 0 4 8 12 16 20 24 28

0 Q4 03 1,?. 1.6 2 CoreA,_ialAverageExposure,Core _al Average Rela_ve Po_r G_O/st

3.1-18

Figure 3.1-13 Equilibrium Cycle Core Data Summary at End of Cycle

-- ,,

Cycle Exposure, MWd/st 7818CycleEnergy,MWD" 1259960 "Number of Full Power Days 321CoreAverage Void Fraction .479

f m , ,,,,

CoreFlow, Mlb/hr 1.151Maximum Channel Peaking 1.242

J ..... Core AxialPower Peak 1.61MaximumRAPLHGR .918

.... MaximumCPRRAT .687,,

Hot Excess Reactivity,% 0.0Cold ShutdownMargin 116 _

N- Numberof3 inchincrementsthatthecontrolbladeis withdrawnfromfullyinserted

28 28

24 ..........................!............................._................!.............i...............22 2220 2O18 ..................._...............................,.................._................... 18

-.¢12 '_ 12

10 _ 108 8

4 4

2 2

0 Q4 (38 12 1.6 2 0 4 8 12 16 20 24 28

COreA_ialAverageRelalive CoreAxialAverageExposure,Pover GWD/st

3.1-19

3.2 LITHIUN TARGETRODDESIGNANDPERFORMANCE

This section describes basic design details of the target rod and reports the

results of perf_ormance analyses for one-cycle exposure. The performance

, assessment includes: thermal analysis, pressure/stress analysis, normal

permeation leakage, failed rod leakage, and tritium distribution in and leakage

from the reactor coolant system. A computer program TEP2was used to perform theanalyses. This code combines all the calculational functions and introduces a

new approach to estimating failed rod leakage (see 3.2.7.1). Details of this

computer code are given in Appendix C, "T2P2: A computer Program for

EstimatingTritium Target Performanceand Tritium EnvironmentalSource Terms".

3.2.1 Design Approach

Reviewof the light water reactorlithiumtarget developedby the TritiumTarget

Development Projecti (TTDP) and its performanceduring normal and off-normal

operatingconditions indicatedthat it is readilyadaptablefor applicationin

the ABWR. Thus, the target rod design propesed is identicalto the reference

design developed by the TTDP except for minor dimensionalchanges required to

interfacewith the standardABWR fuel bundledesign and eliminationof the outer

permeationbarrier.

As a result of the extensive informationavailablefrom the TTDP the approach

used to establishthe ABWR tritium productioncapabilitywas:

• Scale the referenceTTDP targetrod design to fit the standardABWR

fuel bundle consistentwith the design criteria and requirements

establishedby the TTDP2.

• Select a Lie enrichment and target rod placement appropriate to

achievegoal tritiumproductionwith minimum impacton the ABWR core

design.

3.2-1

• For the selected core design determine the Li e n,a rate, the target

rod heat generation rate, etc. necessary to assess target rod

performance and the .,_nual tritium production capability.

• Confirm that contract quantitiesof tritium are produced and that

the targetdesignselectedmeets the performancerequirementsduring

normal operations and transient conditions based on test and

analysisresults from the TTDP.

3.2.2 Reference Design

The target rod, as illustrated in Figures 3.2-I and 3.2-2, consists of

cylindrical,annular LiAlO2 pellets surrounded by a nickel-platedZircaloy-4

getter to absorb and retain tritiumduring irradiation,thus maintaininga low

tritium partial pressure in the free gas space. An inner zirconium liner,

located in the central hole of the annular pellets, assists in absorption of

tritium in the getter by chemicallycracking3H20released from the pellets to

tritiumgas. For conveniencein assembly,the pellets,getters,and liners are

packaged into 12.5-inchlong units called "pencils". There are twelve pencils

in the 163-inch long ABWR target rod.

The pencils are contained in a Type 316 stainless steel cladding tube. An

aluminidecoating is appliedto the inner surfaceof the claddingto provide a

barrier to permeationof tritium into the reactorcoolant.

Table 3.2-I summarizes the key target rod design parameters for the ABWR pre-

conceptualdesign and the TTDP referencedesign.

As indicatedin Table 3.2-I the ABWR targetrod is larger in diameter and has an

aluminizedbarriercoatingon the insidesurfaceonly. The ratio of diameterto

wall thickness is essentiallythe same in both designs. The ABWR rod has a

smaller void volume ratio which results in a higher gas pressure for the same

GVR. However, the clad hoop stress at EOL is stillwell below the unirradiated

yield strength.

3.2-2

Figure 3.2-1 ABWR Tritium Target Rod Cutaway

GETTER(Nickel-Plated

UPPER GETTER Zlrcaloy(Nickel-Plated ALUMINIZED

CLADDINGTOP END Zlr©aloy) TABS BOTTOM

PLUG ENOPLUG

|4_

SPRINGGETTER DISK INNER LINER(Nickel-Plated (Zirconium)

GETTER DISKZircaloy) CERAMIC TARGET (Nickel-Plaled

PELLETS (LIAIO2) Zlrcaloy)IN PENCIL

Figure 3.2-2 ABWR Tritium Target Rod

Table 3.2-1. Target Rod Design Parametersii

_ PARAMETER .... ' ABWR_ NPRREFERENCE_!

Outside Dia. (in.) ..... 0.483 0.371

Barrier Coatin_l I.D. Onlj/ I.D. & O.D.

O.D./(2 x Wall) 8.1 8.2

Rod Void Vol./LiAI02 Vol. 0.8 1.3

Average GVRat EOL 79* 83

Average LHGR(kW/ft.) 0.9 0.4

Average Clad Temp. (°F) 544 610i i llll i

Peak LOCAClad Temp. (°F) <600 1,700

* - In-reactor test data to GVR= 116

The key differencebetweenthe two designs is the peak clad temperatureduring

the design basis loss of coolantaccident (LOCA). This was the limitingfactor

in the NPR referencedesign whereas the ABWR target rod isn't subject to any

significanttemperaturetransients.

3.2.3 Target Rod Performance-Normal Operation

Table 3.2-2 summarizesthe key operating parametersfor the average and peak

target rods in the ABWR core during normal operation. The values in this table

were generated by the T2P2 computer program.

The TTDP in-reactortests operated at clad temperatures-50 °F higher than the

ABWR target rods. However,the n-a reactionrates in the ABWR rods are higher

than in the TTDP test rods. The net result is the ABWR rods have lower clad and

gettertemperaturesthanthe in-reactortestsand comparablepellettemperatures.

The end-of-life(EOL)gas-volumeratio (GVR)for the ABWR rods is well withinthe

range of the in-reactortest data.

3.2-5

The EOLhelium pressure and clad stress in the average ABWRrod are comparable

to the WC-1 test (1,900 vs 2,300 psi and 13.6 vs 9.3 kst). The clad stress in

the peak ABWRrod is higher but still well below the unirradiated yield strength

for 316 stainless steel. Further, the yield strength of the cladding will

increase substantially during irradiation providing additional margin.

Table 3.2-2. Target Rod Operating Parametersi i ,, i

PARAMETER AVERAGE ROD PEAK RODi i i

Clad O.D. Temp. (°F) 542 543i

Getter Temp. (°F) 579 588,,

Pellet I.D. Temp. (°F) 679 712,,,,

EOL AverageGVR . 79 101

EOL Helium Pressure(psi) 2,300. 3,000

EOL Clad Hoop Stress (ksi) . 9.3 14.0

EOL (H+T)/Zr in Getter . 0.3 0.38

Cum. Tritium Perm. (Ci) 0.07 0.09

EOL Tritium Inventory(Ci) 11,400 15,600, ,,

Finally,the getter loading at EOL, adjustedfor hydrogen ingressbased on in-

reactortest experience,is well below the design specificationlimit of 0.7 and

comparableto that observed in the WC-I test.

As indicatedin Section3.2.1 the ABWR targetrod was designed to remainwithin

the criteriaestablishedby the TTDP. The above summaryconfirmsthat this goal

was achievedand furtherthat the normaloperatingconditionsfor the ABWR target

rods are within the experiencebase establishedby the TTDP in-reactortests.

3.2.5 Off-Normaland AccidentConditions

The TTDP conductedextensiveanalysis and testingto determinethe responseof

the getter-barriertarget design to light water reactortransientand accident

conditions. While the TTDP analyseswere based on a 1,250MWe pressurizedwater

3.2-6

reactor(PWR),the followingfundamentalconclusionsrelatedto target response

also apply to the ABWR:

• Because of its low heat generation rate the target rod does not

experience significant temperature increases during over power

transients. For example,in PWR rod ejection accidents3 the target

internaltemperaturewas estimatedto increaseonly 30 °F.

• During a loss of coolantaccident(LOCA)where the core is uncovered

the target will be heated by thermalradiationfrom the surrounding

hot fuel rods.

• The getter-barriertarget can readily be designed to withstand a

LOCA, includingcore uncovery,without clad breach or bulgingto an

extent that would effect coolability.

• All other off-normal,transientand accidentconditionsaddressedin

a typical PWR SAR were much less limiting in terms of target

integritythan the design basis LOCA.

Based on the resultsof the TTDP the only transientor accidentconditionsthat

posed a concernto the integrityof the targetcladdingwere those which involved

core uncovery. Howeveras long as the core remainedcoveredthe target rods did

not overheatbecauseof their low heat generationrate. Since theABWRcoredoes

not uncover during the largest LOCA, there is no concern for target integrity

during off-normaland accidentconditions.

3.2.6 Multi-CycleOperation

As demonstrated in Section 3.1 the ABWR can produce the contract quantity of

tritium by dischargingall the target elementseach year. Since the available

in-reactorperformancedata is limitedto _ I yr exposure,the full core annual

discharge was selected as the reference case. This provides a highly

3.2-7

conservativedesign approach that could be implementedwith the first core if

desired.

The one-third replacement,three year exposure mode more fully utilizes the

burnup capabilitiesof the ABWR fuel and would be a much more economicaltritium

production cycle. A preliminaryassessmentof the potentialto achieve three

cycle exposurelevels in the target rod indicatesthat it should be feasible.

Further,generatingthe in-reactorperformancedata necessaryto supportmulti-

cycle target operationcould readilybeaccomplished using lead test assemblies

in the ABWR after startup.

3.2.7 Impactof Tritium Productionon Plant Operationsand Effluents

The effectsof lithiumtargets on the nuclearcharacteristicsof the ABWR core

were discussedin Section3.1. This sectionaddressesthe loss of tritiumfrom

target rods to the coolantsystem,the buildupof tritiumin the coolantsystem,

and the losses of tritium to environmentalpathways from the coolant. The

tritium is assumedto be chemicallycombinedwith oxygenas T20 or HTO molecules

in the coolant. With this assumption,all calculationsof the tritiumbehavior

are straightforward. The assumptionis easily justified , since all hydrogen

present as gas in the reactor coolant system is recombined with oxygen

continuouslyin the recombiner portion of the off-gas treatment system. The

water produced in the recombinationreaction is re-injectedinto the reactor

coolant system. Since the tritium behaves chemically like normal hydrogen

(protium), there is likely to be only a very small fraction of tritium as

elementaltritium in the coolant.

3.2.7.1 Bases for TritiumTarget Releaseto Coolant

The lithium targets loose tritium through the stainless steel cladding by

diffusionrelatedprocesses. This normal escapemechanismis controlledby the

partialpressureof tritium in the targetrod. There is also the possibilityof

a defect (hole) in the target claddingleadingto depressurizationof the rod.

3.2-8

Normal Permeation Losses

The vastmajority (99.99percent)of the tritiumproducedis containedwithin the

LiAl02pellets and the getter. The partialpressureof free tritiumin the gas

space of the target rod during normal operationis only _ 10.4atm. This partial

pressure is determined by the dynamic balance among the process involved in

releaseof tritiumfrom the pellets,its absorptionin the getter and liner and

permeation through the clad. These processeshave been modeled in the TKTARI

code describedin the literature4. This code, and an evolutionaryupgradeused

in this study, combinetheory and validationexperimentsto give a mechanistic

calculationfor the permeationreleaseof the tritium from the targets during

irradiation. For the design and irradiationparametersof the ABWR target the

calculatedtritiumpermeation(T2P2code) is O.07Ciover the 274 day irradiation

period per rod or 244Ci for the 3,488 rods in the core.

Losses from Failed Rods

No cladding failures occurred during the in-reactor testing and no failure

mechanismswere identifiedother than fabricationflaws,externaldamage and gas

pressure inducedstresswhich is readilyaccommodatedin the design. There was

no mechanical interactionor chemical attack among the cladding, getter and

pellets. However,the possibilityof claddingbreachesin the targetrodscannot

be dismissed

While there is no failureexperiencefor getter-barriertargetrods considerable

data exist for the failureof fuel rods in light water reactors (LWRs). The clad

failure rate for LWR fuel is generallyacceptedto be in the range of 1/10,000

rods (ReferenceI). The target rod failuremodel used in the Phase IA studies

assumed a clad failurefrequencybased on LWR fuel elementexperience and that

50% of the tritium inventory in the failed rod would be released. The 50%

release impliesthat all claddingfailuresexist from beginningof life and that

the capacityfor retainingtritiumis determinedby the solubilityof tritium in

the LiAl02. This was consideredto be very conservativeon the basis that the

most likely target rod failureswould occur late in life when almost all the

3.2-9

tritium would be tied up in the zirconium getter and LiAI02 pellets and not

available for release. However, because the tritium permeation from intact

target rods is so low, the release from failed target rods using this model

dominatesthe tritium sourceterm.

To overcome this limitationa more mechanisticand realistictarget rod

failuremodel was developedas summarizedin the attacheddescriptionof the T2P2

code. The overalleffectof the new model is to delay the targetrod failureto

an exposuredeterminedby the Weibulldistributionfunctionand the frequencyof

reactorshutdown-restartcycles. This reducesthe impactof clad failurebecause

only tritiumreleasedfrom the LiAlO2 after clad failureescapesto the coolant.

This revised failuremodel predicts that 130 Ci of tritium are released to the

coolantper refuelingcycle which for the referencecycle is one year.

Total Iritium Losses to the ReactorCoolant System

The total sourceterm per cycle can be summarizedas 244+130= 374Ci/year. Since

both the normal p__rmeationand the releasefrom failed target rods are expected

to be very gradualthis sourcetermhas beenlinerizedas 1.4 Ci/dayduring power

operationbased on 274 full power days per cycle (i.e 374/274 = 1.4).

3.2.7.2 TritiumDistributionin the Coolant and HVAC Systems

The water inventoryintowhich the tritiumcan mix during operation includesin

the reactorvessel,the feed and condensatesystem,the condensatestoragetank,

the rad waste system and the spent fuel pool. This total mass of coolant

(I.538E+07pounds) is partiallyredistributedduring refuelingto include the

dryer-separatorpool. For simplicityof calculationthe tritium is assumed to

be "well-mixed"in this total inventory. The only differences in water loss

during power operationand refuelingperiodsare no steam leakageto the turbine

buildingduring refuelingand the spent fuel pool exposes 300m2 to evaporation

and the dryer separator pool, which is filled during refueling, essentially

doubles this evaporationarea.

3.2-10

Evaporationfrom the pools is assumed to be controlled by the mass transfer

coefficientfor evaporation(K.)and the pool-to-reactorbuilding air humidity

(watervapor concentrations).Using a literaturevalue of 3000cm/hrfor K,B with

the pool at tOO°F,the evaporationlosseswere calculatedfor a reactorbuilding

refueling platformHVAC flow of 35,000cfmwith the flow assumedto be well-mixed

in the buildingspace. The inlet HVAC air was assumedto be 70°Fat 50 percent

relative humidity. The resulting evaporationrates were 16,4001b/dayduring

power operationsand 31,3001b/dayduring refueling.

The ABWR is designedas a closed systemwith no routineliquiddischargesto the

environment.However, based on experienceit is expected that _ 10,000gal/day

(83,0001b/day) of makeup water will be required to compensate for pool

evaporationand miscellaneoussteam leaks. Based on the pool evaporationrate

indicatedabove the 10,000gal/daymakeup rate impliesthat _ 66,600 Ib/day are

lost through steam leakage, primarily in the turbine building,during power

operation. The loss of water from the discharge of solidifiedrad waste is

negligiblein the overall tritiumbalance.

To accountfor the gradualbuildupof tritiumin the coolantsystem,an unsteady-

state mass balanceof tritiumin the systemwas modeled in differentialequation

form and integratedfor severalcycles of power operationand refueling. In the

model it was assumed that the tritium is present as water and no separation

occursduringpool evaporation.The overallloss rate includedradioactivedecay

but was dominated by the water loss terms. Results of the analysis show the

tritiumconcentrationin the primarycoolant increasingduring the power cycle

and decreasingduring the refuelingoutageas a resultof dilutionwith the water

from the spent fuel _nd separatorpools. However,the tritiumconcentrationin

the primarycoolantreachesa repeatinglevel in two to three refuelingcycles.

The effectivehalf life of tritium in the reactor coolant system is only 0.28

years. The average coolantinventoryover each cycle reachesa steady level of

154Ci or 1.0E-O5Ci/Ib. The modelalso provides a tabulationof tritium source

terms for worker exposure and site dose calculations.

3.2-11

3.2.8 Summaryof Airborne Tritium Releases

Tritium losses to the HVAC system are again assumed to be well-mixed in the

building space to determine the building airborne concentrations. These

concentrationsare also time-averagedover both the power and refuelingperiods

of each cycle. Table 3.2-3 summarizesthe pertinentcalculatedresults needed

for the dose calculations.

Table 3.2-3. Summaryof Tritium Concentrations and Source Terms*

;' J i t i i, iiJl ,

Liquid Phase 154 Ci averagecycle inventory_ ..... 1.0E-05Ci/Ib avera_lecycle concentration

Gas Phase Turbine Building (TB)347 Ci lost**/cycle@ power0 Ci lost/cyclerefueling1.5E-07#Ci/cc in TB @ powerO. #Ci/cc in TB refueling

Reactor Building (RB)31 Ci lost/cycle@ power24 Ci lost/cyclerefueling7.8E-08#Ci/cc in RB @ power

.... ].gE-07pCilcc in RB refuelin9

* - Steady operation (>3 cycles)with 274 days of effectivefull powerfollowed by 91 days refueling.

** - "Ci lost" are time-integratedvaluesthat do not includedecay afterexitingthe stack.

3.2.9 References

I. PNL-8142 "Tritium Target DevelopmentProject Executive Summary iopical

Report", W. J. Apley, September1992.

2. WHC-SP-0840"Topical Report" NPLWR TritiumTarget Design", J. W. Weber,

September 1992.

3.2-12

3. RSA-010 "Assessment of 10% Core Performance During Reactivity Insertion

Accidents Through End-of-Life", B. E. Schmidt, et al, Pacific Northwest

Laboratories, October 1991.

4. WHC-SP-0684 "TKTARI: A Computer Code for Predicting Tritium Target Rod

Performance,"D. R. Wilson, 1991.

5. "Handbook of Chemical Property EstimationMethods, "W. J. Lyman, W. F.

Reehl, and D. H. Roseblatt,1992.

3.2-13

3.3 TRITIUM TARGETFABRICATIONANDRECOVERYFACILITY REQUIREMENTS

The tritium target fabrication methods are based upon work conducted under the

Light Water Reactor Tritium Target Development Program (LWRTTDP) (Reference 1).

The tritium recovery (extraction and purification) methods are derived from

requirements established during the LWRTTDP, but are also based on existing

facility capabilities and purification requirements of the Replacement Tritium

Facility (RTF) at the Savannah River Site (SRS).

3.3.1 Target Rod Fabrication

ABWR targetrod fabricationinvolves:1) producingthe LiAl02pellets,cladding,

getter and liner components,2) assemblingthe pellets, liner and getter tubes

into pencils,3) loadingthe pencilsintothe claddingand 4) completingthe rod

final assemblyand end cap welding. A simplifiedprocessflow diagramfor target

rod fabricationis shown in Figure 3.3-I.

The Light-WaterReactorTritium Target DevelopmentProject (LWR TTDP) utilized

commercialvendorsto fabricatenickel-platedgetter tubes,aluminizedstainless

steel claddingassembliesand the LiAlO2 pelletsutilizedin the TTDP in-reactor

tests. While additional work remains to qualify vendors for full scale

production,a high level of confidencewas establishedthat commercialvendors

could be qualifiedto fabricateall of the target rod components. Discussions

with the FabricationTask Managerfor the LWR TTDP supportthis approach. Itwas

also identifiedthat portions of the fabricationprocessand componentdesigns

may be classified. This was consideredin the evaluationof alternativesand

appeared to be no differentthan many other DOE or DoD contractswhich involve

classifiedinformation.Therefore,the referencecasefor fabricationof tritium

target rods is commercialvendor productionof all target rod components.

The pellet fabrication process begins with enriched lithium carbonate and

aluminum oxide which are blendedand spray dried. After drying, the powder is

calcined in a furnace and a dry binder added using a blender. Multicavity

hydraulicpressespress the pelletsand a belt-feedcontinuousfurnace sinters

Dry Binder -_

LithiumCarbonate

Blend and SprayDry _. Calcine _, Blender _ Hydraulic Press

Aluminum ._

Commercial Vendors

Stainless Steel Tubes --7 Getter -----]

Helium Pellet [_ Pellet Inspection,Atmosphere ." Tube Loading ,_ ,. _ ",, _ _ Sampling, and _" Sintering Furnace

Drying _uoassemolv

Target Assembly

!'

Helium Leak Test

Final Target Rod _. Welding _. and Weld _. Store .-'_ Transfer to FuelAssembly Radiograph Assembly Area

Figure 3.3-1. TargetFabrication Process Flow Diagram

the pellets. The pellets are inspected and sampled and centerless ground to

final size. The pellets are then assembled with getter tubes and liners into

pencils and stored until required for loadtng into the cladding tube assemblies.

No development or qualification for large scale production of pellets was

initiated during the LWR TTDP. A stngle batch of lithium aluminate was

synthesized for the test program. Two methods of pellet fabrication were

utilized and both seemedadequate. Isostattc compaction and uniaxial pressing

were both used. A commercial vendor was used to fabricate approximately 630

pellets by tsostatic compaction.

The barrier-coated 316 stainless steel cladding Is fabricated by a pack

aluminizing process in which a mixture of aluminum alloy powder, alumina powder,

and an ammoniumchloride activator is blended in a predetermined mixture. The

prepared pack is loaded into a clad tube which has the lower end cap already

welded in place. A group of packed clad tubes is then placed in a retort and

heated under conditions that produce the desired barrier coating. Eddy current

and air-gaugeinspectiontechniquesare used to assureuniformityof the barrier

coating. It was determinedin the early phases of the LWR TTDP to pursue the

barriercoating developmentutilizinga commercialcoatingcontractor. It was

assumedthat the commercialvendor would impose a productionorientedapproach

during the developmentthat would enhance future scale up to productionscale.

Initialdevelopmentof the barriercoatingprocesswas performedon 48 inchtubes

due to furnace availability. Further development on full length tubes was

initiated. Several coatingparameterexperimentswere conductedand eventually

coatingparameterswere establishedthat resultedin homogeneousmicrostructures

and uniform thicknessalong the full interiorlength of the tubes.

The nickel-platedzircaloy (NPZ) getter tubes are fabricatedusing commercial

electro-platingtechnologyfollowedby vacuumannealing. The NPZ conceptrelies

on the electroplated nickel layer, highly permeable to tritium and highly

resistantto oxidation,to diffusionbond to the Zircaloy,providinga means of

entry for the tritium into the Zircaloy storage media. A commercial tubing

supplier, capable of making acceptable getter tubing by redraw of existing

reactor grade Zircaloy-4 fuel cladding tube stock, was found through the

competitiveprocurementprocessduring the LWR TTDP.

3.3-3

The inner liner is a roll-formed tubular shape of O.O04-inch zirconium metal.

Metal shapes of this geometry are routinely fabricated in industry and a

commercial vendor should be readily available.

The final rod assembly consists of loading subassemblies in air on a work bench.

Loaded rods are inserted manually into a portable evacuation/drying magazine.

Each magazineholds a one-day throughput. The internal atmosphere is replaced

with dry, high purity helium. Each magazine is allowed to soak until the

components are within acceptable limits for moisture and oxygen. The magazine

would then be connected to a welding enclosure and flushed with high purity

helium. The springs and top end caps are inserted into the tubes and the end

caps welded. Whenall the end welds are complete the magazine is uncoupled from

the welding enclosure and the rods removed and placed in the helium leak-test

chamber. Following final inspection,the rods are placed in containers and

transportedto a bundleassemblyareawhere targetrods, fuel rods and associated

hardware are assembledinto the fuel bundle for shipment to the reactor.

With commercialfabricationof the targetcomponents,the onlytarget fabrication

operationsto be performed as part of the target rod assembly are:

- Target rod componentinspectionand storage

- Pencil assembly

- Rod loading

- Helium back fill

- End closure assembly and welding

- Final rod inspectionand packaging.

Based on the design developed for the LWNPR support facilities,approximately

15,000 target rods per year can be fabricated in an industrial-type,single

story, 15x30m building. The ABWR would require only 25 percent of this

throughput. However, the need to handle 4m long rods horizontallyrequires

approximately 10m of working length. Allowing space for component storage,

inspection,helium back fill and welding equipment, a total of approximately

300m2 should be more than adequate for target rod assembly.

3.3-4

The equipmentrequired for target rod assembly includes:

- Two 5m long assemblytables

- Five evacuation/drying magazines

- Helium fill station equippedfor pump-downand back-t'ill

- Helium glove box with two automatedGTAWmachines- Helium leak testing and weld radiography equipment

- Miscellaneous inspection tools and instruments.

The targetrodassemblyoperationis a normalindustrialactivity.Thereareno

specialsafety,licensing,workerhealthand safety,effluentor waste issues

associatedwith this activity.The onlywastesgeneratedthatare potentially

regulatedunderthe ResourceConservationandRecoveryAct (RCRA)arethe spent

radiographyfilm processihgchemicals. Thesewasteswould be collectedand

packagedas requiredfor disposalby a commercialche_'calwastecontractor.

Safeguardsand securityassociatedwith targetfabricationis relatedto the

enrichedlithiumpelletsandpotentially,protectionof anyclassifieddocuments

related to the fabricationprocess. The protectionand accountability

requirementsassociatedwith enrichedlithiumwhich is classifiedas "Other

NuclearMaterial"(Category4, AttractivenessLevelE) underDOE Order5633.3A

(Reference2), arelessstringentthanfissilematerialandaresimilarto those

associatedwithpreciousmetalsin the DOE system.

Finalassemblyand inspectionof the targetrods couldbe handledin several

ways. Thereappearto be threeprudentoptions.PhaseIAdiscussedassemblyof

the componentsin the MOX Fuel FabricationFacility. This is stillvalidand

expansionspaceis beingretainedinthecurrentlayoutof theplant.Additional

optionsconsideredin thisphaseincludeutilizinga commercialfuel fabricator

to performthe targetassemblyor performanceof thistaskat a DOE facility.

Theremay be politicaladvantagesfordecouplingthe tritiumtargetfabrication

fromtheMOX fabrication.Thisis particularlytrueif theMOX Plantwas placed

under internationalinspectionas part of a bilateraldisarmamentagreement.

This couldeasilybe accomplishedutilizinga qualifiedcommercialvendorand

3.3-5

implementing the security measuresrequired to protect classified information.

There are manyfavorable aspects of targe' assemblyutilizing a commercial fuel

fabricator. These fabricatorsare intimatelyfamiliarwith fabricationof

assembliesof similarphysicaldimensionand components.The fabricationwill

requiresimilarweldingand inspectioncurrentlysupportedin existingfuel

fabricationfacilities.QualityprogramsincorporatingNQA-Irequirementswill

be very similar. Physicalsecurityof thematerialswill alsobe similar. If

thedesignof the targetrod is classified,additionalmeasureswillneedto be

implementedatthecommercialfacility.However,the incrementalcostassociated

with classificationshouldbe less if the targetrods are fabricatedat a site

where personnel and physical security measureswere already being imposed.

It is also feasibleto performthe targetrod assemblyat an existingDOE site

in a modifiedor newfacility.Thefactthatthereare no radioactivematerials

involvedimposestherequirementthatthefabricationfacilitiesbe cleanrather

thanexistingfuelfabricationfacilitiesthathavebeenshutdownunlesstheycan

be adequatelydecontaminated.An existingDOE sitewouldhavethe advantageof

in-placeinfrastructurefor safeguardsand security. A decontaminatedfuel

fabricationfacilitywouldalsohavethe advantageof experiencedpersonneland

technologyrequiredfor the target assembly,along with stringentquality

programs.

All informationacquiredsupportsthe use of commercialvendors for the

productionof componentsfor the tritiumtargetrods. Italsoappearsfeasible

to utilizea commercialfuelvendoror DOE fuelfabricationcapabilitiesforthe

assemblyof completetargetrods. It is notedthatsomeadditionaldevelopment

work is requiredpriorto productionscalefabricationof LWR TritiumTarget

Rods. Theseare addressedin Section6.3.

3.3.2 TritiumRecoveryFacilityRequirements

The phaseIA report(Section6.6) (Reference3) describedthe processrequired

to extract tritium from the ABWR target rods and performthe necessary

purificationand isotopicseparationsto obtainthedesiredproductquality.It

also includeda descriptionof the supportsystemsneededto minimizeworker

3.3-6

exposure and environmental releases. Except for the head end extraction

operation, all of these process capabilities are included in or provtded for the

Replacement Tritium Facility (RTF) whtch has just been completed at the SRSand

is currently undergoing operability testing.

It was determined during meetings with SRSpersonnel not to introduce the new

extraction off-gas into Building 232-H because tts useful life cannot be

guaranteed for the life of the PuDisposition project. The 232-H factlity is 36

years old and Pu Disposition introduces a need in about ten years for a life of

about 30 years. The level of upgrades and modifications required to retain the

232-H facility product extraction capabilities would not be cost effective.

Previous studies have been conducted addressing addition of extraction capability

to the RTF. The "Replacement Extraction and Purification Facility (REPF)--

Building 231-H" Functional Performance Requirements was issued February 16, 1990

(Reference 4). This project proposed extensively greater purification

capabilitiesthan will be proposed for the Pu Destructionmodification. It is

noted that some of these other modificationswill still be needed for the RTF to

operatewithout the supportfunctionsnow providedby the 232-H facility. These

are being addressedin anotherrequirementsdocumentfor the ReplacementTritium

PurificationFacility (RTPF) (Reference5). The RTPF is currentlyunfundedand

studiesare being conductedto justifyupgradesfor extendingthe use of Building

232-H to the year 2050 for off-gastreatmentwithoutextraction (resultingin a

minimal source term within the old facility).

If the RTPF were funded to provide replacementpurificationcapabilityfor the

aging 232-H facility,it would be most prudentto add the LWR Target Extraction

Facilityto it (resultingessentiallyin the original REPF but with LWR rather

than HWR extractioncapability). Consideringthe unfundedstatus for the RTPF,

there are three viableoptions: I) a greenfieldfacilityat the reactor site to

extract and purify tritium gas prior to gas shipment to the SRS tritium

facilities,2) the identicalgreenfield facility adjacent to the SRS tritium

facilities for extraction and purification,or 3) a new extraction hot cell

locatedadjacentto the RTF with supplementalpurificationcapabilityinstalled

within the expansion area of the RTF. The cost of options I and 2 would be

3.3-7

similar for any location and for any of the LWRoptions being considered. This

cost will not be specificallyestimatedsince it will be comparablefor all LWR

options. It is anticipatedthat the greenfieldfacilitywould be downsizedfrom

the REPF which was sized to supportRTF and 100% of goal tritium productionin

1989 (capitalcost ,,$200M).

;" Performingthe extractionof the SRS requiresonly a few annual spent fuel cask

shipmentsof irradiatedtargetsand the materialshipped is less attractivefor

theft or diversion than a purified tritium product. See Section 4.4.7 for

discussionsregardingthe transportationoptionsconsidered.

Since the gas purificationand isotopicseparationsystemsplus their associated

supportsystems(blanketgas,purge gas, etc.) representthe majorityof the cost

of a LWR target processingfacility,the recommendedapproach is to add a target

extraction facility (with a shielded hot cell) to the north end of the RTF.

Pretreatmentcapabilitywill also be added in the expansionarea in the north end

of RTF, becausethe existingRTF diffuserand TCAP capabilitiesare specifically

for processing returns without protium. The RTF TCAP currently separates

deuteriumfrom tritium. It is anticipatedthat th_ additionalsupportservices

required for this approachwill tax the existing RTF support systems to their

originaldesign capacitieswhich were sized for future expansion.

Figure 3.3-2 shows the 200-H Area Tritium Facilitiesand the proposed location

of the ExtractionFacilitynorth of the 233-H building. The extractionfacility

is sited adjacent to stairwellS-2 such that a doorway from the grade level

landingof the stairwellcan enter directly into the shieldedoperatingarea so

that separatesecurityaccesscontrolis not requiredfor personnelaccessto and

from the extractionfacility. A new railroad spur is indicatedto providerail

car access for standard spent fuel shipping casks which will be used for

transportof the target rods. Figures3.3-3 and 3.3-4 show the plan view and

elevationof the targetextractionfacilitythatwould be locatedadjacentto the

expansionarea on the north sideof the RTF. As indicated,the buildingconsists

06 a 18x20xgmhigh reinforcedconcretestructureenclosinga 10x10xS.Smhigh hot

cell and a rail car bay for receiptof standardLWR spent fuel casks. The rail

3.3-8

233-IH

Figure 3.3-2 Tritium Facilities

3.3-9

i 20.0m i

10.0m i

/ Railroad SteelFrame4 ft.ReinforcedConcrete

Ful'll_C¢

Steel Wall

AirLock

StorageHolesE

0

06Shidded

M_" MagazineStorage DoorC>

TransferCanUnloading FurnaceMagazineLoading

Figure 3.3-3. Tritium Extraction Cell Plan View

i 20.0 m iI00Ton Crime

HVACSpace

f:: 5TonCrane

C)

Railroad EBay _

Hot Cell

/ ShieldPlugs Gallery

.........GRADE............_,........!

100SteelStorageTubes6"Dia.x 14'Ling

Cask TransferTmmel

Figure 3.3-4. Tritium Extraction Cell -Elevation

car bay ts equippedwtth a 100 ton crane and a below-grade cask transfer tunnel

that provides access to the top loadtng spent fuel casks through the hot cellfloor.

The hot cell ts dtvtded tnto an atr-ce11 and a nitrogen-cell by a metal

partition. The partition ts less than full height wtth a closed top makingthe

furnace portion of the cell a sealed nitrogen-filled box. Thts permits the tn-

cell crane to operate over the nitrogen boxwhich ts equippedwith access ports

for furnace removal. The nitrogen box can also be equippedwtth a wall or

ceiling mounted light duty electro-mechanical manipulator tf needed. The

partition also contains an air lock for transfer of furnace magazinescontainingtarget rods to and from the nitrogen blanketed furnace cell.

The process gas, blanket gas and the argon purge streamswould be connectedto

the corresponding RTFsupport systemsfor processing andtritium recovery. The

extraction furnace system includes valves, gas receiver tanks and vacuumpumps

which would be located in the RTFexpansionarea as shownin Figure 3.3-5 along

with the required pretreatment systemsto removeimpurities from the furnace off-

gas stream prior to introduction into the existing RTF processes. The furnace

off-gas in the receiver tanks will be sampledutilizing a 200 ft capillary tube

to the Hass-Specglovebox in room010. This samplewill be for accountability

along with verification that the extraction is complete. It is anticipated that

the small amountof impurities (i.e. protium and hydrocarbons) introduced into

the RTFfrom the samplebefore pretreatment will be tolerable.

Pretreatment capabilities will be addedto the RTFin the expansionarea at the

north end of the facility. Pretreatment is required to removeprotium from the

product gas. This will require gas separation and isotopic separation. The

process gas from the extraction receiver/hold tank is first pumpedthrough a

heated uraniumbedto decomposethe oxides of tritium andhydrogen. Gasfrom the

decomposeris cooled and pumpedthrough the primary uranium hydride bed to

collect the hydrogen isotopes. The effluent gases (primarily helium and

impurities) are sent through a palladium/silver sacrificial bed to remove

impurities that could damagethe diffuser, then pumpedthrough a multistage

3.3-12

I l_nt Northt I New Pretreatment Area _ ([_

IL 1 _ __,_ New Extraction Hot Cell

0441

Figure 3.3-5. New Pretreatment Area in RTF

diffuser to recover the remaining hydrogen isotopes. The gases that flow through

the diffuser tubes, primarily helium, are collected in one of two tanks which can

also be sampled to determine tritium content. These gases are sent to the RTF

off-gas cleanup system where they are further treated, if necessary, or sent to

stack discharge. The hydrogen isotopes, which diffuse through the tubes, are

collected dtrectly on the secondary urantum hydride bed. Hydrogen isotopes are

released by heating the hydride beds, then pumped to the thermal cycling

absorption process (TCAP) feed tank. The TCAPcycles the gas between two beds

of palladium-coated kteselguhr (Pd/K). One bed is heated to drive off the gases

while the other is cooled to promote gas absorption. The palladium-coated

kteselguhr preferentially retains the tritium and thus separates it from the

other hydrogen isotopes (prottum). Whenthe pressure in the TCAPproduct tanks

reaches the desired level, the gas is assayed and pumpedto the TCAP Product

storage (2-500 ltter) or TCAP Feed storage (2-500 liter) beds in the TCAP

glovebox (T105-300-1) located in Room017 of RTF. The raffinate material is sent

to the RTF off-gas cleanup system for removal of residual tritium and is then

stack discharged.

Room ventilationfor the operatingareas of the new extractionfacilitywill be

provided with the building. It is planned that the RTF stack (and effluent

monitoring system) will be utilized for all gaseous discharges from the

extractionfacility. The nitrogenblanketedfurnacecell will be connectedto

the existing stripper system in RTF. The air portion of the hot cell will be

HEPA filtered (to remove potential particulate materials introduced during

storage or transport of the irradiatedtargets) prior to discharge to the RTF

stack.

The air cell includes storage holes for 125 percent of the annual reactor

dischargeof targetrods, an unloadingstationfor the transfercans used to ship

target rods from the complexto SRS and a furnacemagazineloadingstation. The

nitrogencell includestwo vacuum furnaces,their associatedsupportequipment

and the transfermechanism for loadingand unloadingthe furnaces.

3.3-14

The spent target rod after extraction Is expected to retain approximately -50

curies of tritium in the LtAI02 pellets and have a gammadose rate in the 100

R/hr range at one foot. Interim storage for spent target rods at SR$or other

provisions (e.g. return to the reactor site) wtll probably be required if the

spent target rods are disposed of as core componentstn the federal spent fuel

repository. Althoughdtsposal of the spenttarget Podsin the federal repository

ts permissible (Reference 6), this maynot be as cost effective as dtsposal asa low-level waste at the SavannahRiver Stte (see Section 4.4.7 for further

discussion).

3.3.3 Tritium Cost Considerations

TaraetFabrication

TargetfabricationcostsinthePhaseIAreportwerebasedon estimatesdeveloped

duringthe LWR TTDP. No new issueshave been identifiedwhichwouldwarrant

changingthis estimate. Target assemblyat a vendor locationshould be

comparableto the estimatedcostto assemblethem in the MOX Plant.

TritIum Recoyery

Tritiumrecoveryhasseveraloptions.Acommon optionforall LWRfacilitiesis

a greenfieldextractionand purificationfacilitythat shouldbe in the $200M

capitalcostrangesimilarto theREPFprojectthatwas developedby SRS. This

greenfieldfacilitycouldbe locatedat the reactorcomplexor at SRS nearRTF.

One differencebetweenthe two locationswould be transportof the tritium

productversusirradiatedtargetrodsto the SRS.

Our referencecase is the additionof an extractionhot cellnextto the RTFat

SRS as proposedand costedin the PhaseIA report. An additionalmodification

to the RTF is requiredin our PhaseIC referencecasethatwas not includedin

the PhaseIA reportestimates.This is the "PretreatmentFacility"locatedin

the expansionareaof the RTF alongthe northend of the building. This will

includetwogloveboxesthatcontainuraniumhydridebeds,a diffuserand a TCAP

3.3-15

wtth capacity to process 38 mtllJon curies of tritium per year. Based on

discussions with SRS personnel, the estimate for this modification to the

existing RTF ts in the $15-2SM range. This was not tncluded in the ortgtnal

Phase 1A estimate.

3.3.4 References

1. PNL-8142 "Tritium Target Development Project Executive Summary Toptcal

Report", W. J. Apley. September 1992.

i

2. DOE Order 5633.3A "Control and Accountability of Nuclear Materials",

Department of Energy, February 12, 1993.

3. NEDO-32292, "Stuay of Pu Consumption in Advanced Light Water Reactors,

Evaluation of GE Advanced Boiling Water Reactor Plants", GE Nuclear

Energy,May 13, 1993.

4. WSRC-RP-gO-147"FunctionalPerformanceRequirementsfor CurrentAppraisal

of Cost, Project S-3312, Replacement Extraction and Purification

Facility(REPF)(U)Building231-H", H. W. Harmon,January 30, 1990.

5. WSRC-TR-91-442 "Functional Performance Requirements for Replacement

Tritium PurificationFacility(RTPF)(U) Building 231-H, ProjectS-4568",

H. W. Harmon, July 15, 1991.

6. IOCFR Part 961 "Standard Contract for Disposal of Spent Nuclear Fuel

and/or High-level RadioactiveWaste", Department of Energy, January I,

1992.

3.3-16

3.4 ABWR-PDR PLANT OPERATION FORTRITIUM PRODUCTION

3.4.1 Introductioni

During phase la of this project (Reference i), the option ofconvert the reactor system to produce tritium was evaluated.In was shown that tritium production goals can be met by theuse of four lithium aluminate target rods per assembly in astandard UO 2 fueled ABWR core. Rod design, fabrication andproduction considerations were described along with anestimate of the impact on plant operations and effluents.Support facilities associated with tritium production alsowere described. The results of these investigations clearlyshowed that goal quantities of tritum can be produced withthe ABWR.

The current evaluations in Phase ic address the use of a

mixed oxide (MOX) reactor core for the purpose of disposingof weapons plutonium while also inserting Lithium targetrods for the production of tritium. This section discussesthe impact on plant operations associated with the tritiumproduction mission.

3.4.2 Basis for Review

Two goals, to be satisfied concurrently, have beenestablished by DOE. I) the disposal of I00 MT of Plutoniumpreviously used in the US weapons program, within 25 yearsand 2) the production of Tritium on an "as-needed" basis tosupport long term DOD needs.

The planned operation consists of two distinct operatingcycles: a plutonium disposal cycle in which MOX bundles areburned in approximately 18 month operating cyclesconsistent with the power production and disposalobjectives; and a tritium production cycle in which Tritiumtargets are exposed in a MOX fueled core for about 1 year.Steam produced is used to produce electricity which is sold

to a local utility for transmission. Estimates of keyexposure parameters are summarized in Table 3_4-i.

During the Plutonium Disposal Cycle, about 232 MOX bundlesare removed to the reactor pool (and fresh fuel is loaded)approximately every 18 months. The removed fuel is allowed

to cool in the reactor pool for about 1 year before beingtransported to a high level waste depository or a temporarystorage facility.

3.4-1

When operation of a unit is redirected for tritiumproduction, a full core offload and reload is required.Partially spent MOX fuel is stored in the fuel storage poolwhile a new core with targets is inspected, channneled and

loaded into the core. To provide room in the spent fuelpool, transport of previously discharged MOX fuel may berequired in preparation for the Tritium Cycle.

Following the Tritium Cycle, the entire core is offloaded tothe spent fuel pool and the previously operating MOX core isreloaded into the core to continue the plutonium disposalCycle. Tritium targets are removed from the exposed bundlesin the reactor pool and transported to an offsite extractionfacility. The residual MOX bundles (without the targets)are reconstituted for use in a future cycle.

3.4.3 Preparation for Tritium Production

3.4.3.1 Fuel Inspection and Handling

Section 3.2.3 of reference 2 discussed the potential needfor underwater inspection of fresh MOX assemblies. Althoughfresh MOX fuel may be mildly radioactive, no specialhandling is anticipated. To avoid design modification toaccomodate underwater inspection in the new fuel vault,fresh MOX fuel will be inspected and channeled in the samemanner as current fuel designs. If occupational exposurebecomes a concern with this approach, underwater inspectionin the spent fuel pool may be persued as an option.

Addition of the target rods to the MOX fuel bundles is notexpected to require special handling beyond that used forthe recovery of the exposed targets. Therefore nomodifications or other considerations are needed.

3.4.3.2 Fuel Handling

Initiation of the tritium production cycle requires a fullcore offload of the previously operating core and storageuntil the tritium production cycle has been completed.

The need for full core offloading places an added burden onthe fuel storage pool in the ABWR-PDR design. The currentdesign provides for storage of about 270% core loading(Reference 3) or about 5 cycles of operation withoutaccomodating the Tritium production mission. To provide fora full MOX core offload and Targeted core onload, less spacewill be available for spent MOX storage. The current ABWR-PDR design will accomodate about 2 cycles of spent MOX.Therefore, in preparation for Tritium production, it is

likely that offsite shipment of at least 3 cyc__ _ of spentMOX to the waste disposal facility will be required. It isassumed that such facilities are available and that shipmentcan be accomplished within the six month preparation period.

No impact of the target rods is expected on the overall fuelloading or refueling process since the targets are containedin the fuel assemblies. However, due to the need for fullcore offload of the tritiated core and reload with the

previously operating MOX core, care must be exercised toassure that fuel assemblies are well tracked and locations

verified. Enhanced tracking of fuel bundles and theirexposures, with and without targets, has a higher importanceto meet the concurrent goals of disposal and tritiumproduction. Equipment such as LASERTKAC may be beneficialfor this ABWR-PDR use. Such equipment is not currrentlyincluded in the ABWR certified design, but could beconsidered for future development.

3.4.3.3 Fuel Pool Cooling

The ABWR fuel pool cooling and cleanup system (FPCC) isdesigned to accomodate heat removal from 35% an offloaded

core 21 days after shutdown and the spent fuel from 4previously offloaded cycles (Reference 3). If higher heatloads exist, the RHR system may be aligned to providecooling assistance.

Following the initiation of a Tritium production cycle, thefull offloade MOX core will require additional cooling.Therefore, during the some portion of the Tritium cycle, oneloop of the RHR system will be aligned in Fuel Pool assistmode. No plant modifications are needed to accomplish this.Operability of this RHR loop in other operating modes (suchas Low Pressure Co£e Flooding or Suppression Pool Cooling)will rely upon manual realignment, as required.

3.4.3.4 Other Design Considerations

The MOX core design results in a higher fast flux and lowerthermal flux from the MOX core in comparison with uraniumfueled cores. Initiation of tritium production decreasesthe effect. Because of these differences, certain systemsare potentially affected.

Neutron Monitoring/Calibration, Process computer heatbalances, SLC boron requirements and the radiationmonitoring systems were reviewed to determine if other

design considerations would be affected by initiation of thetritum production cycle. No significant differences were

3.4-3

identified which would either require a design modificationor increase operating/maintenance costs.

3.4.4 Operation during Tritium Production Cycle

3.4.4.1 Operating Limits and Power/flow map impact

Section 2.8.1.2 of reference 1 indicated that a higherminimum recirculation flow (35% Speed / 45% Flow) is neededwith the MOX core to avoid the instability region.Analysis of changes resulting from initiation of tritiumproduction have not been conducted, but slight changeJ arenot expected to result in operational difficulties.Specific procedural changes reflecting changes in theoperating limits, rod patterns and the power/flow map willbe needed and appropriate operator training provided priorto plant restart for the tritium production cycle.

3.4.4.2 Occupational Exposure and Routine Offsite Releases

Table 8.2 of reference 1 estimated the average annualexposures from the ABWR at about I00 Manrem (not includingoperation of the fuel fabrication facility). Tritiumleakage was not explicitly included. Initiation of thetritium production cycle would be expected to increase thedose recieved at power and in the turbine building due topotential for required breathing apparatus in some areas andthe resultant inefficiency associated with the maintenancework. In addition refueling activity may be slightly lessefficient due to airborne tritium above the refueling pool.If the exposures recieved in these activities are increasedby 10% due to the inefficiencies, the total annual exposurewould be increased by about 2 man-rem. Therefore, asignificant increase is not expected due to initiation ofthe tritium production cycle.

Additional discussion of the routine offsite releases from

steam leaks and evolution from the refueling pool isdiscussed in Section 3.2.

3.4.4.3 Abnormal Occurances

Target Rod Failure Impact

Evaluation of tritium rod performance (Section 3.2)indicates that the internal rod pressure due to helium gasis appoximatley 2300 psig following one exposure cycle of273.75 days. The stress from such pressure on the targetrod, however is a small fraction (approximately 10%) of theyield stress for the rod under operating conditions. It can

3.4-4

therefore be concluded that failure is not likely. However,if failure were to occur, the impact of such sudden failurewould be minor since the accumulated helium has no

significant impact on the operation of the reactor or plantsystems. On the other hand, the sudden failure of a targetrod could have an adverse impact on plant performance if itleads to a significant release of tritium comparable withhydogen water chemistry (HWC) upsets or resin intrusionevents.

HWC operation injects approximately 2 ppm of hydrogencontinuously in order to suppress the radiolysis occuring inthe reactor core. For the ABWR with full feedwater flow of

17.1 x 106 Ib/hr, about 34 ib/hr of hydrogen would beinjected into the core. Since, based on the evaluations inSection 3.2, the amount of tritium available for release

from a single rod very small (about 6x10 -9 grams), it isclear that failure of a single rod would lead toinsignificant amount of hydrogen gas.

Decay heat Implications

Figure 2.7-19 of reference 1 shows the expected Decay heatfrom an equilibrium MOX core. Because exposure from thetritiated core would be on a fresh core, decay heat levelsfollowing any accident or transient would be significantlylower and, because of the nuclear differences, the MOX core

would have a slightly lower decay heat than a comparablesize UO 2 core. Because of these considerations, shutdowncooling and other shutdown safety issues are less severe inthe tritiated core and during the plutonium disposal cycle.

3.4.5 Termination of the Tritium Production Cycle

Fuel Handling Issues

No modifications are considered to be necessary for thehandling of the tritiated core.

Target Rod Removal

Tooling and procedures to remove exposed target rods fromexposed fuel assemblies within the ABWR-PDR pool need to bedeveloped and tested.

Fuel Reconstitution

Following the tritium production cycle, the remaining MOXbundles exposed along with the lithium targets will have

significant life remaining for additional burnup. Tooling,

3.4-5

procedures and methods to permit reconstitution of thesebundles for additional exposure have not been developed andtested at this time. However, no impediments to developmentof such methods have been identified.

3.4.6 References

I. "Study of Pu Consumption in Advanced light WaterReactors", GE Nuclear Energy, NEDO-32292, May 13, 1993.

2. "Study of Pu Consumption in Advanced light WaterReactors, Compilation of Phase Ib Reports", GE NuclearEnergy, NEDO-32293, September 15, 1993.

3. "Safety Analysis Report, Advanced Boiling WaterReactor", 23A6100 Rev i, Amendment 31.

Table 3.4-1

Key Parameter Study

...... Tri titan Plutonium

Production DisposalCycle Cycle

Number of Reactors 1 6

Cycle length (EFPD) 273.75 392.2 .....MOX Bundles 872 232

Dis charged/cyc Ie

Plant Capacity Factor 75% .... 75%Discharge Exposure N/A 37081 ......(MWd/MT)

100MT Plutonium Disposal N/A i9.3Time (years)Tritium Production ii,595 N/A(Ci/rod) - 3488 rods

t

4.0 INFRASTRUCTURE AND DEPLOYMENT

4.1 MOX FABRICATION INFRASTRUCTURE

4.1.1 JAPANESE FACILITIES

OVERALL SUMMARY:

1. A plant to reprocess LWR fuel, privately funded by Utilities (75%) and

Industry (25%), is being located at Rokkasho in northeast Japan.

2. A privately funded MOX plant that will utilize the reprocessed Pu, to serve

both BWR and PWR reactors, is still in the planning stage.

During preliminary meetings of the industry/utility group it was decided that

while the plant will be highly automated, it will be built to allow non-remote

maintenance if required.

3. Reprocessing of Japanese LWR fuel is expected to be carried out in Europe

until about 2002, after which both reprocessing and MOX fuel fabrication will be

shifted to Japan.

The utilization of Plutonium in thermal reactors has not yet approached a

practical stage in Japan. At present Japan has only a single small-scale MOX fuel

fabrication facility for LWR's. This facility is under the administration of the Power

Reactor and Nuclear Fuel Development Corporation (PNC) and is located at the

Tokai Mura site north east of Japan. A new large scale MOX fuel fabrication facility

will be required before full scale utilization of plutonium in thermal reactors can

progress.

Over the past year, the nuclear community in Japan has been studying the

requirements for a MOX fuel faciity to to realize the full commercialization of

Plutonium in LWR's. The MOX fuel facility has been targeted for start-up in the early

years after the turn of the century. The plant capacity is estimated to be around

100 MT MOX per year. The location of the facility has yet to be decided but it is

anticipated that the plant would be located near the reprocessing site in Rokkasho

Mura in northern Japan.

Early in December a visit was made to Japan to obtain more detailed

information on the status of the Japanese MOX program and in particular the MOX

fuel fabrication facility study. The details of this visit are covered in the next section.

4.1 .I-I

Visit with Hitachi, December 1, 1993:

Japanese MQX Pv0grarn"

The Japanese program, in the MOX area was described by Dr. M. Oguma. Hitachi is

involved in a number of areas of the Japanese MOX program, including MOX fuel

designs for the reactor, MOX factory design and licensing, MOX shipping container

designs and as part-owner of the reprocessing facility. A facility to reprocess LWR

spent fuel has been approved and is being built at Rokkasho in northeast Japan. This

facility, Japan Nuclear Fufels Ltd (JNFL), is a consortium with 75% funding from

utilities and 25% funding by the industry. A MOX factory, to serve both BWRs and

PWRs, is still in the planning stage. At the time of this writing, this factory was also

to be privately funded, however, there were indications that there was some

consideration given to government ownership of this factory.

The reprocessing factory was expected to ship a master blend of 50% UO2- 50%

PuO2 to the MOX factory.

The proposed schedule for Pu utilization is shown schematically in Figure 1. Until the

year 2002, LWR fuel will be reprocessed in Europe, in a number installations

(including British and French plants in the future). Beyond this date, it is expected

that the entire LWR spent fuel load would be reprocessed in Japan.

All facilities will be subject to IAEA inspection standards.

Question and Answer Session:

The following answers were given to specific questions during the Q&A/DiscussionSession:

Process:

The MOX factory will employ a standard mechanically blended MOX pellet

fabrication, similar to the refrerence process identified by GE for the disposition study.

The input will be the 50-50 master blend and the output will be full (BWR or PWR or

Fast reactor) bundles.

4.1 .I-2

4j_ ~ 70 MT/yr

Mox JapanTonnage European _ .f Reprocessing

1 I I I I I1993 95 97 99 01 03 05 07 09

Fig 1 Schematic of Proposed Japanese Plan for LWRMOX (reprocessed)Fuel Uti:ization

4.1 .I-3

Throughput:

The peak MOX factory throughput is pegged at 100 MT of MOX per year. The

average capacity of the MOX plant will be 70MT of MOX per year which should

satisfy the LWR recycling need.

MOX Fuel Fabrication Experience, Fuel Additives:

Only PNC has practical MOX fuel fabrication experience. The Hitachi/Toshiba

designs are based on input from PNC and from visiting European installations. They

have no experience with poison additives to MOX fuel other than Dysprosium used in

small quantities for power shaping in the MOX fuel bundles irradiated in Tsruga 1 unit

in the 1980s. The poison additives proposed for the U.S. Pu disposition project are

used for reactivity control rather than power shaping.

Level of Automation:

The Japanese had visited MOX facilities in Europe and thought the

Belgonuclaire plant had the maximum throughput although they were not the most

highly automated. Hitachi and Toshiba with JNF are the MOX factory designers and

stated that even though they started with a very high level of remote operation, they

have since adopted a strategy that will allow "non-remote" maintenance.

Transportation:

Transportation from Europe, in the form of completed assemblies for LWRs or

for the fast reactor, has already been designed. It is expected that the MOX factory

will be co-located with the reprocessing factory, however, this decision has not been

made. If they are not co-located, they do not see any problem in shipping the master

blend MOX powder as this is "routinely" done in Europe. Hitachi and Toshiba are

presently designing the transportation casks for shipment of completed MOX bundles

from the MOX factory to the reactor sites. In most instances, they expect this

transport to be carried out over water routes rather than inland routes. Casks for sea

transportation are already available and those for road transport are under design.

4.1.I-4

These designs might be applicable to a US program.

Safeguards:

Hitachi admitted that although they are in the process of preliminary design for

the MOX factory, they have not completely incorporated Safeguards requirements into

the design process. All the plants will be subject to IAEA inspection and standards.

Hitachi allowed that only PNC had real experience in material accountability and

safeguards and that any new technology for implementing safeguards would have tocome from PNC.

Licensing of the MOX factory:

Government (MITI) is the process of establishing the licensing criteria. It is

likely to be modeled after European standards and could possibly be used as a basis

for U.S. certification.

Licensing of the Reactor plan for MOX:

Hitachi is doing MOX fuel designs. Only partial MOX core loads (up to a third

of the core) are envisioned. They do not see any problems nor the need for lead tests.

The nuclear parameters were verified with MOX fuel test assemblies in the Tsuruga 1

reactor in 1986. These assemblies were discharged in 1990 with 25000 MWD/MT.

The MOX assemblies are not located at peak power positions and therefore can attain

the same exposure as the urania bundles over a longer period of time. The bundles

will be full MOX designs and not "island" MOX designs.

It was Hitachi's view that for Pu fractions up to 5%, no additional data was

needed. Beyond about 10%, Hitchi felt that there were sufficient changes to the fuel

properties (e.g., solidus/liquidus temperature)and that more detailed assessmentswould be needed.

4.1.1-5

Waste:

The waste from the MOX factory will be mainly a result of the scrap during

fabrication. It is to be divided into "clean" scrap and "dirty" scrap. By definition,

"clean" scrap will be sent to an appropriate entry point in the fabrication process to be

recycled and blended into the process flow. "Dirty" scrap which cannot be.blended

into the process flow and which quite possibly requires solutioning with resulting

aqueous streams, will be returned to the reprocessing facility. Thus, they do not feel

there will be any real waste stream from the MOX factory itself and all the waste will

be handled as part of the waste stream in the reprocessing facility.

Hitachi did not provide any cost estimates. The schedule called for about 8

years to complete the MOX factory from project initiation to initial bundle production.

Of this, 3 years were for design and licensing, 3 years for construction and 2 years

were reserved for test runs. They anticipate only 1 shift operation while in production.

Visit with Toshiba, December 3, 1993:

The role of Toshiba is parallel to Hitachi in the Japanese MOX program. A visit was

arranged with Toshiba for two reasons. First, in Japan, it would not be considered

proper to visit only one of these two industry participants, even if all the relevant

information would have already been obtained from the first meeting. Second, it

served to confirm the details obtained at the first meeting.

Japanese MOX Program:

The Japanese MOX program was described by Toshiba, including their role, as

a fuel MOX designer, fuel factory designer, and transportation cask designer. They,

like Hitachi, are investors in the reprocessing facility and expect to invest in the MOX

factory. They clarified that some of the Pu from reprocessing will be used by PNC for

the fast reactor. During 1991-92 time period, concepts for the MOX plants were

examined and they expected a MOX plant to produce about 70 MT of MOX per year

on the average, 45 MT for BWRs and 25 MT for PWRs. The MOX factory design is

still in the planning stage. Most of the work is being carried out by Toshiba, Hitachi

and JNF. They have completed a layout, equipment requirements and process flowsheets.

4.1 .I-6

All the questions asked of Hitachi was also asked of Toshiba. Toshiba

confirmed Hitachi's answers in each of the areas. Specific areas where Toshiba was

able to provide some additional information are noted below:

Hot-Cell vs. Glove Box Type Lines:

To accommodate both BWR and PWR bundles, it would be necessary to •

change the enrichment frequently. GE pointed out that this could lead to considerable

down-time in cleaning out the line and complying with material accountability. Better

availability could be obtained with a combination of glove-box line type with a hot-cell

type set up where the entire contents of the hot-cell could be swapped with a standby

unit, to minimize downtime and enhance accountability. They (Toshiba) have not

considered these areas in detail.

Cost and Schedule:

Toshiba indicated the cost of the MOX plant to be in the neighborhood of US$1

billion. It was noted that this was considerably higher than the 90% finished Siemens

plant which was reported to have cost DM900 million. It was also noted that by using

existing facilities, principally structures which meet required licensing standards, the

cost could be lower. The large discrepancy between the German plant and the

Japanese estimate was left unresolved. No O & M estimates have been made.

Visit to JNF, December 3, 1993:

JNF is a major urania fuel fabricator in Japan. JNF is a principal investor in the

reprocessing facility and is expected to play a major role in designing, building and

operating the MOX factory.

The meeting was attended by M. Petski and T. Shigeto from GE. Principal

contacts at JNF were K. Murota, Chief Engineer, K Kumoro, Senior Staff Member and

T. Ishikawa, Section Manager, Business Dept. T. Ishikawa was identified as the

principal member of the Japan MOX planning team from JNF. Ishikawa also had spent

I0 years on assignment with the IAEA in Vienna.

JNF is located in Kurihama, approximately one hour south of Tokyo, on the

entrance to the Tokyo Bay. JNF is a joint venture operation of the General Electric Co

(40%), Hitachi (30%) and Toshiba (30%). The venture was first established in 1967

4.1 .I-7

and fuel fabrication started in 1970 under a technology agreement with GE. JNF

produces fuel to designs provided by the three companies (shareholders).

The plant capacity is 850 MT U02 per year. Present fuel production is 570 MT

U02, equivalent to 3000 bundles per year.

The operations are 75% automated from powder receiving through bundle

assembly. There are two fabrication facilities on site. Both are licensed to handle 5%

enriched fuel. The main plant has a capacity of 580 MT. The sub plant has a capacity of

270MT and is used for fabrication of the Gadolinia fuel.

JNF fabricates its own spacers, tie plates and other small components. UO2

powder is procured from GE in Wilmington NC, France, England and Sumitomo in

Japan. Zircoloy tubing is supplied through GE Wilmington, Sumitomo and Kobe in

Japan.

JNF has assigned several people to work on the Japan MOX planning team.

This team meets on a weekly basis to integrate plant and equipment design inputs.

Preliminary plant and equipment designs have been proposed and are under further

team study and refinement. Preliminary cost numbers for the facility are in the one

billion U.S. dollar range. There seems to be some feeling that this number may be off

because of the uncertainty of the cost of safeguards. The PNC Tokai facility is

considered by the IAEA as the model for safeguards standards. The team feels that

integrating these practices into a manufacturing facility would greatly increase cost

projections. These safeguard differences still need to be resolved. No one could

identify a completion date for these studies but an early 1994 completion date was

estimated.

MOX LEAD USE ASSEMBLY TESTING:

Toshiba presently has the lead in the fabrication of fuel assemblies for

insertion into a Toshiba designed reactor. Plans call for loading of MOX bundles

equivilant to 1/4 of core size. The MOX fuel for this program will be fabricated at BN

in Belgium and the Bundles assembled at FBFC in France and shipped to Japan. The

target reactor for the fuel assemblies is Reactor 3 at the First Fukushima site. JNF

will handle the administration of the fabrication contract, qualification of Belgium and

French facilities, and will supply all UO2 rods and bundle hardware. This activity

will take place over the next 3 to 5 years. JNF has also been directly involved with

Hitachi and Toshiba in MOX studies at BN in Belgium, COGEMA in France and

BNFL in England. From the manufacturing perspective, JNF has, over the past

4.1 .I-8

several years, participated with Hitach, Toshiba and others in MOX bundle transport

container design and fabrication, and equipment design related to automated rod

inspection ,bundle assembly and welding equipment for MOX fuel manufacture.

4.1 .I-9

4.1.2 BNFL Experience and Facilities for MOX Fuel Fabrication

4.1.2.1 Experience Base

A. Fuel Cycle Capabilities

BNFL and its predecessor, the AEA, have been designing and manufacturing fuel for nuclear

reactors, enriching uranium, transporting fuel, reprocessing spent fuel and managing the waste

products for over 40 years. BNFL was created out of the former Production Group of the

AEA in 1971 and became a public limited company in 1984. In 1990, BNFL Inc. was

formed as a wholly owned US subsidiary of BNF.

In the UK, the company's principal business is the provision of the complete cycle of nuclear

fuel services as the supplier to the UK nuclear utilities. In addition, the company has

extensive experience in the provision of fuel for many types of reactor systems including

Magnox and AGR, water reactors (PWR, BWR and SGHWR) and fast reactors (PFR). This

includes uranium metal fuel for the Magnox reactors and ceramic oxide fuel for the other

reactor types, including MOX for the fast reactors.

In the US, BNFL Inc. has grown rapidly with a number of design and engineering projects,

largely in the DOE waste management area, but also with a number of commercial clients. In

addition, BNFL Inc. has performed project management for the transport of nuclear materials

from US sites to the UK for BNF.

B. History of Involvement in MOX Fuel

BNFL and its predecessor, the AEA, firstbegan manufacturingMOX fuels in the early 1960s

when about 3 tonnes HM was produced for a wide variety of reactor systems including PWR,

BWR, and gas- cooled reactors. BNFL and AEA recognized the importance of fuel

homogeneity and indeed, all the fuel performed well in reactor, thus demonstrating the

feasibility of using plutonium fuels in LWRs.

From 1970 until 1988, BNFL and AEA produced over 18 tonnes HM of fast reactor MOX

fuel containing plutonium at enrichments of up to 33 %. A very close working relationship

with fuel cycle development has been maintained between the fast reactor fuel fabrication workin BNFL and in AEA.

In addition to the large scale manufacturing experience gainedto support the Fast Reactor

4.1.2-1

Program, AEA at Windscale also provided a service for the manufacture of experimental fuel

for the project. The experimental fuel fabrication facilities at Windscale were operated for

almost 20 years and during that time were used for fabricating a variety of different uranium-

plutonium containing fuels (oxides and carbides), a range of pellet sizes, various designs of

fuel pins, different cladding materials and irradiation rigs. Following manufacture, fuel was

transported from the Windscale/Sellafield site to the reactor facility. In addition, BNFL has

had experience of transporting plutonium in nitrate and oxide form to customers

internationally.

C. Plans to Provide Thermal MOX Fuel Capability

As a major reprocessor BNFL is committed to the effective utilization of its customers

reprocessing products. The provision of a MOX fuel supply capability is an essential part of

that strategy.

In March 1990, BNFL and AEA established a formal collaboration agreement in the Thermal

MOX Fuels business area. As a demonstration of commitment to providing a secure and

reliable thermal MOX fuel service to its customers, BNFL in collaboration with AEA Fuel

Services invested in the design and construction of a small-scale MOX fuel production facility

(MDF) which is approaching completion at Sellafield. The experience gained in the

experimental fuel fabrication facilities at Windscale is directly relevant to the technology and

production of modern thermal reactor MOX fuels and will be fully utilized in the fabrication of

MOX fuel in the MDF. In particular, the plant incorporates the BNFL short binderless route

process which produces a highly homogenous MOX powder fuel for fuel pellet production.

This facility will be capable of producing up to 8 tonnes HM per year as PWR MOX fuelassemblies from the end of 1993.

BNFL is further developing its interests in the MOX fuel market through the construction of a

commercial scale plant in the UK known as the Sellafield MOX Plant (SMP). The BNFL

board has recently approved the next stages in the design and construction of SMP which is to

be operational by the end of 1997. The plant will incorporate expertise from MDF and other

fuel fabrication facilities. The plant will be designed to be highly reliable with low

maintenance requirements and low in-line stocks supplemented by off-line secure storage

facilities. The operation, containment, shielding, and maintenance of equipment will be

designed to achieve stringent BNFL targets for radiation exposure levels. The plant capacity

will be around 120t HM/year based on market assessments for the requirements of MOX fuels

towards the end of the 1990s. To achieve the tight program time scales required to ensure the

4.1.2-2

completion of SMP by end 1997, BNFL has drawn together a multi-discipline Task Force

within their offices at Risley, Warrington. This Task Force currently comprises a substantial

Team with expertise drawn from a number of recently completed major projects at Sellafield,

(including input from MDF commissioning experience), and fuel fabrication expertise from

Springfields. At this time the project team is more than 200 strong and is expected to reach

approximately 300 by early 1994. The task force includes project-management, all main

design groups, safety assessors and project services. In addition specialist input from, for

example, shielding and criticality assessors is being extensively used. The SMP process is

now fully established and is supported by an extensive development program. Building layouts

are complete with design progressing for commencement of construction and procurement

programs early in 1994. The first issue of the associated safety case has recently been

submitted to the regulators.

It is the expertisewithin this task force linked with BNFL Inc.'s US personnel which has been

used in the compilation of this report.

4.1.2.2 Description of the Sellafield MOX Plant

To support initiatives for MOX fuel supply, BNFL is currently embarking on the construction of

a large scale Mixed Oxide Fuel Fabrication plant on the Sellafield Site. This will give increased

capacity beyond that available from the MOX Demonstration Facility (MDF) and enable return of

plutonia to customers as usable MOX thermal reactor fuel. This plant will be called the Sellafield

MOX Plant (SMP).

The purpose of the plant will be to receive feed powders (plutonia and urania), convert them to

MOX pellets and subsequently manufacture fuel rods and assemblies. The plant is being designed

to manufacture a nominal 120 tonne heavy metal per year of MOX fuel assemblies. SMP will be

capable of producing a wide range of products in the form of PWR and BWR rods and

assemblies.

The manufacture of fuel pellets will be achieved using a process developed by BNFL known as

the short binderless route. Unlike conventional binderless granulation techniques where the urania

and plutonia powders are ball-milled for a considerable period to give the required properties, the

short binderless route utilizes a high energy attrition mill. This permits short cycle times of about

40 minutes compared with 4-8 hours for a conventional mill but still ensures that 100% of the

material is blended. Upon production of a homogeneous powder mixture, the conventional pre-

compaction and granulation has been replaced by a new process which utilizes a spheroidiser.

4.1.2-3

This agglomerates the powder in the presence of small quantities of die lubricant to give free-

flowing granules for the press feed.

The short binderless route offers a number of advantages:

a) Homogeneous pellet structure.

b) Good dissolution properties.

e) Good fission gas retention

d) Small hold-up in plant.

e) Fully contained process

f) Simple process which is easy to maintain and operate.

The urania feed material will be produced by the BNFL owned Integrated Dry Route (IDR)

process at BNFL's Springfields site. Over 14,000 tonnes of urania powder have been made by

this process which has been licensed in the US and France. The mechanical, physical and nuclear

properties of IDR material control to a large extent the properties of the MOX fuel when mixed

with typically 5-8% of plutonia. Following pellet production, the fuel manufacturing route

follows conventional practice. The majority of processes within SMP are fully automated to

reduce the dose uptake to the workforce.

The plant will incorporate campaign operations for producing pellets of a single enrichment for

loading into fuel cans. Campaign sizes will vary and could range from 0.5 to 10 tonnes. The

process plant is being designed so as to minimize loss of throughput due to enrichment and

campaign changes.

4.1.2.3 Process Description

The overall SMP process flow sheet is given in Figure 4.2.1-1. To support the

effective management of the design and subsequent installation of Mechanical,

Electrical and Instrumentation equipment, SMP has been divided into ten plant

areas.. Areas 100-600 correspond to the main stages in the process, and 700-

900 cover the service functions. Area 000 covers the general equipment within

the building.

4.1.2-4

These plant areas, as currently defined, are as follows:

Area 000-General Building Equipment

Area 100 -Powder Receipt

Area 200 -Powder Processing

Area 300-PelletingArea 400-Rod Fabrication

Area 500-Fuel Assembly

Area 600 -Sampling/Effluents and Contaminated ResiduesArea 700-Ventilation

Area 800 -Control, Electrical and Instrumentation (CE&I) SystemsArea 900-Services

Powder Receipt - Planl; ArQa 100

The following powder feeds are supplied to the SMP plant:

Plutonia (THORP, MAGNOX COGEMA)

Urania granules.

Urania granules combined with Zinc stearate lubricant.

CONPOR, this is a proprietary material used to control the porosity of the

sintered pellets.

Zinc stearate which is required as a lubricant in the spheroidisers.

The above powders are supplied in containers and are placed in an appropriate

storage on receipt. From storage, each powder container is transferred into a

glove box, weighed and then opened. The powders are transferred,

pneumatically, to a feed hopper when required for powder production.

Powder Processing - Plant Area 200

Powder production comprises 3 major stages. These are:

a) Milling

b) Blending

c) Spheridising

4.1.2-5

I'

The powders (plutonia and urania) are weighed, in the correct proportions, and

transferred into the feed hopper. From here they are fed, under gravity into an

attrition mill to enable the milling of the powder to take place.

The CONPOR powder is added to the process to control the porosity of the

sintered pellets. Zinc stearate powder is added to the homogenization process to

act as a lubricant in the attrition mill, spheridiser and blender.

The milled batch is fed, under gravity, to a blender. Three mill batches are

required for each blender operation. From the blender the blended material is fed,

via a screw conveyor, to a further attrition mill and then to a spheroidiser whe,_ it

is granulated. The granulated material is fed via a screw feeder to the press

hopper.

The common powder feed system supplies two parallel process towers from

milling to pellet pressing inclusive.

Pelletiing - Plant Area 300

Pellet Pressing

Granulated material is fed from the press feed hopper into the press where it is

pressed. Pellets produced are inspected, loaded onto boats and transferred to the

sintering furnace.

Sintering Furnace

The pellet boats are fed into one of four sintering furnaces via gas locks. The

boats are pushed through the furnace at a constant rate. Within the furnace the

boats pass through a series of zones. These are:

a) reducing zone

b) sintering zone

c) cooling zone

The sintered pellets are removed from the furnace still on the boats, via a gas lock

and are transferred to the grinding process glove box.

4.1.2-6

Pellet Grinding

The pellets are unloaded from the boats, passed through one or two grinders

where they are ground to size. The pellets are then passed through a series of

brushes where residual dust particles are removed. The pellets are checked for

size, inspected for surface defects and then loaded onto trays. Pellets may be

removed during inspection for sample analysis if required. Sampling systems will

be employed with use being made of the existing THORP sample transport

systems. The pellet trays are conveyed to a sintered pellet store.

Rod Fabrication - Plant Are_ 400

Sintered Pellet Store and Stacking Systems

The trays are weighed into storage and placed on shelves awaiting results of

analysis. The trays are weighed out of storage. The pellets are removed from

these trays where they are formed into stacks on stack trays containing thecorrect enrichments for the fuel rods.

Rod Filling, Welding and Inspection

At one of the two rod fill and weld stations, empty rods are purged with helium

prior to the pellets being inserted into the rods from the stacking system. The

filled rods are moved to a welding station where a cap is TIG welded onto the end.

The fuel rod is pressurized via a drogue hole which is subsequently welded. The

fuel rod is then passed through a series of inspection stages:

a) leak test

b) mass spectrometer

c) X-Ray inspe,:tion

d) rod scanner

e) identification/contamination inspection

f) geometry inspection

g surface texture inspection

h) straightness inspection

4.1,2-7

Rod Transport

Finishedfuel rods are insertedinto magazinesprior to manufactureof the final fuel assembly.

Fuel rods are received from the rod inspectionstages, identified,aligned in the correct positionand inserted into a magazine,positioned at the insertion machine. The orientation within the

loaded magazinereflects the requiredorientationof the fuel pin arrayin the final fuel assembly.Magazines arestoredin a dedicatedmagazinestore.

FueJAssembly-Plant Area 500

AssemblyProductionand Inspection

The loaded magazines are transported to one of the two fuel assembly lines. Skeletons are

constructed from components on a bench. (Skeletons may also be supplied to SMP pre-constructed.) The magazines areasalignedwiththe skeletonandthe fuel rods are loadedinto theskeleton.

Whenthe fuel rods havebeen installed into the skeleton,bottom andtop nozzles are fitted where

appropriate. Thecompleted fuel assemblyis inspectedinitiallyfor:

a) geometry

b) grid location

c) rod to rod gaps

The assembly is subsequently cleaned, if required.

The final inspection of the fuel assemblycomprises of:

a) cleanliness

b) damage scratches

c) grid and spring condition

d) weight check

The inspected fuel assembly is stored until it is required for despatch in a suitable transportcontainer.

4.1.2-8

PCM Waste Handling- PlantArea600

During normal SMP operations, solid wastes will arise principally in the form of plutonium

contaminated material _CM). These arising can be split into two main streams:

a) process feed waste (mainlyempty plutonia cans)

b) maintenancewaste

The process has been designedto minimizeas far as practical the generation of secondaryPCM.

The PCM wastes are exported inside drums to the site waste treatmentcomplex via the PCM

marshallingroom withinthe THORPcomplex.

Ventilation - PlantArea700

The main functionsof the ventilationsystem are:

a) To assist in providingactivitycontainment.

b) To provide a satisfactoryworkingenvironmentfor personneland equipment.

c) Remove airborneactivity fromdischargeairto ensureemissionsareacceptable.

The overallventilationsystem will comprise a numberof separatesystems designed as a cascade

to ensureairflows from areas of lower contaminationclassificationto areas of potentially highercontamination.

The system for contaminationclassificationis broadlyas follows:

a) C4/C5 Zones: Normallyunmannedareasincludingglove boxes.

b) C3 Zones: Occasionallymanned areassuch as process cells containing gloveboxes and maintenance areas.

c) C2 Zones: Normally mannedareas such as access corridors,change rooms,

operating working areas.

d) C1Zones: Inactive areas includingoffices and change rooms.

The main components of the ventilation system are:

a) C5 Glove box Extract. The glove box system will be at the greatest depression and air

will flow into the glove box from the surrounding C3 cell areas. Each glove box extract will

be filtered and connected to a combined C5 extract duct. This will be filtered and monitored

4.1.2-9

prior to dischargevia the THORPStack.

b) C3 Cell Ventilation,Air will cascade into the C3 cell areasfrom the C2 operatingareas.C3 air extractedfrom the cells (i.e. that which does not cascade into the Glove box C5

System)will be combined in a C3 extract duct. This will be filteredand monitoredprior todischargevia the THORPstack.

c) C2 Operating Area Ventilation.SMP will have a dedicated C2 air supply and extract

systemwhichwill providefiltrationand monitoringof extractedairpriorto dischargeattheTHORProof level.

d) C1 AreaVentilation. C1 areaswill be servedby commercialtypeventilation system.

..ControlandInstrumentation-PlantArea 800

Location of Control

In line with the overallphilosophy to minimize operator dose uptake the majority of the process

plant and services will be controlled, monitored and surveyed remotely from a central control

room. This will be located in the SMP process building. Some fuel assemblyoperations will be

controlledlocally using control stations dedicatedto each operation. The main control functionscarriedout fromthe SMP control roomwillbe:

a) plant productioncontrol

b) fissilematerial accounting

c) monitoring and display of process parametersand alarms

The plant environmentalmonitoring systems (e.g. activity-in-air,ventilation and service supplies)will be controlledand monitoredfromwithinthe mainTHORPcentral control room.

4,1.2-10

Typeof Control

IntegratedAutomation System (IAS) consisting of ProgrammableElectronic Systems (PES) will

be used to carryout the operationalmonitoringandcontrol function of the process. The IAS will

conform to relevant BNFL and InternationalStandardswith respect to integrity levels and willprovidethe following control features:

a) Startup, normalrunningandshutdownof operations

b) Sequence control ofbatch operations

c) Stop and/orreset of individualoperations

d) Authorization of releaseof plant equipment to maintenancemode

The IAS does not containanycomponentsof the engineeredsafety protectionsystems which are

necessary to support the safety case. These are providedin accordance with the requirementsof

the appropriatecompany standardsand utilizefully independentcomponents from the IAS. The

status of systems defined as protection systems as well as the status of plant parameters and

alarmsdesignated as safety relatedare displayedin the SMP controlroom.

4.1.2-11

MassFlowrateAppmxlmately600 kg/day

i I MILL' 50kg i_ MILL0

!

150kg I BLENDER i BLENDER

50kg ' MILL t MILL ORfI!

,,i

, f• 50kg ti PRESS PRESS _Samplest .

I " jlf ....._- 1 _' I _

lt .....I_....i' .... I, I

!GRINDER I GRINDER _.._day

; TRAY(PELLET)STORE

fROD LOADING ROD LOADING

I

L

Figure 4.1.2:1' SMP'Process Flow Diagram

4.1.2-12

4.1.3 COMPARISON OF U.S. AND FOREIGN (UK) MOXFUEL FABRICATION FACILITY REGULATORY

REQUIREMENTS

Although DOE Orders and NRC regulations for handling and processing Special Nuclear

Materials (SNM) exist, as pointed out in Phase 1B studies, there has been no regulatory activity

since mid 1970s that specifically concerns the requirements for constructing and operating a MOX

fuel fabrication facility. In this section, a comparison has been made of the existing U.S.

regulations with the those applied in UK for the Sellafield Mixed Oxide fabrication Plant (SMP), to

identify the key differences between the applicable requirements and their impact on cost, schedule

and development requirements.

The following comparison is based on construettng the current design of SM.Pwithin the U.S. Thecomparisonhas been conducted b)' reviewing the SIvIPdesign criteria or current BblFL standardsagairt,_tthe equivalent US Standards and regulations. Where key differences have b_n highlighted the ',ff'feetofth,_c on cost. program and development requirements has been identified. Inorder to simpli_,"the reviewthe comparison was carried out against the following meas:

a) Safety and Environmental Standardsb) Safeguards Requirementsc) Security Provisionsd) Regulatory Approvals

To assist the reader information has also been provided on the following:

a) Program tim,'tales for the construction of SMP within the UK,b) lnl'rastructurc requirements currently available on BNFL's Sellafield _ite which will be utilized by

the SMP.

c) Current identified development rcquiremet_ for SMP.

4.1.3-1

4.1.3-1 SAFETY & ENVIRONMENTAL

This safety and environmental comparison was carriedout by reviewing the SMP radiological safetyrequirementsand criteria(see Appendix.23)against the comparableUS regulatory requirements. It isrecognized that within the US each plant will have internaloperationallimits. These are more stringentthanthe regulationsin orderto avoidapproachingthe overallregulatorylimits. However. in the absenceof any information on exact location of the plant a review against the operational criteria was notpossible.

I, COMPARISON

The following section is a comparisonof the criteria applicable to SMP (as described in Appendix A)against the comparableUS regulations.

I.A. Occupational Limits

The principle dose target for BNFL Plants is that dose uptake to individuals should not exceed 15mSvy"*with the group average < 5mSvyj. The US Regulatory requirement, translated into a quarterly limitof 12.5 mSv is less restrictive than the current BNFL design target in terms of occupational exposure.

BNFL's _ practice is to operate plants with monthly dose uptake "limits" of .3mSv (whichcompares favorably with the US Regulatory limit).

The Engineering Design Principles which are used by BNFL to ensure compliance with the ALARPconcept should also be adequate to demonstrate the US ALARA concept which is largely similar.

I.B. Accident Conditions

For large accidents the US Regulations limit the public dose at the site boundary to 250roSy. The USRegulators would expect such accidents to have a frequency of occurrence < 10_y_. The BNFLapproach to accidents arising from internally initiated events with an equivalent consequence would beto limit the event frequencyto < 10_y_. The BNFL requirement is therefore more restrictive in thisrespect.

The BNFL criteriaforaccidentsaffectingthepublicalsolimitthefrequencyofsmalleraccidents(<250mSvatsiteboundary),e.g.,effectivedosesbetween0.OlmSvandImSv arelimitedto0.Oly* foraerialdischarges.Similarlimitationsareplacedoneventsleadingtoaccidentalliquideffluentdischarges.TheBNFL requirementis,inthisrespectmore restrictive,byrequiringtheminimizationofaccidentalaerial/liquiddischargesfromotherthanlargeaccidents.TheBNFL designcould,therefore,includemoreprotectivesystemsthanwouldtheUS equivalent,assumingthattherearenofurtherUS Regulations.

The inclusionofa timeaverageddosecriterionintheBNFL designrequirementsistosimplifythedemonstrationthatthesummedfatalrisktoa member ofthepublicfroma siteis< 10C'y_. No

equivalentmortalityrateisgivenfora US site.

The BNFL designrequirementsareclearlymore restrictivethantheUS Regulationsforcontrolofaccidentaldoseuptaketooperators.The annualprobabilityofexceedanceoftheUK StatutorylimitisallocatedatargetofI03 whileeventsleadingtodoses> ISvareallocatedanannualprobabilityofI0_.There is no equivalent in the US Regulation. In addition to these requirements, smaller accidents whichlead to airborne contamination/abnormal dose rates sufficient to cause evacuation are restricted to

..

4.1.3-2

< I0_:'j. The US Regulations stipulate that workforce doses from accidents on a plant will beaccommodated within occupational limits (50mSvy_) for other than large accidents(where the workforceare considered 'disposable'), In BNFL terms, any accident which would reasonablybe expected to occurwithin the lifetime of a plant (say once in 50 years) would be included within the predictive dose uptakeassessment (occupational dose) for the plant, taking into account the potential consequenceand predictedfrequency. However application of the BNFL Engineering Design Safety Principles which wherepracticableare applied to all new plants (althoughnot mandatory) will ensurethat the consequences fromaccidents occurringwith this frequency would be very low.

BNFL plants are also designed to ensure that the potential for criticality events to occur is lessthan 10_ _. The building fabric will be such that the 100 mSv contour is within the building itself.

I.C. External Event

Seismic protection on BNFL plants is determined accordingto the design basis earthquake which, in thecase of SMP is the event with a return period of 10_years (ie once in 10,000 yrs). In the event of a DBEoccurring, seismic qualification of plant and equipment will be designed to ensure that the maximumpossible dose to an off-site person will be less than 5mSv. Plant which do not have the potential to resultin an unmitigated consequence of > 5mSv but > lmSv will be designed to withstand a seismic event witha probability of exceedance of 10_. All other seismic protection is made against ALARP considerations.This is considered equivalent to a design against the US "Safe Shutdown Earthquake" although it shouldbe determined how the UK and US 104yreturn period seismic events compare.

An operating basis earthquake (O.B.E.) is defined as that which would reasonably be expected to occurin the lifetime of the plant (once in 50 years for SMP). No safety related plant, system or structurewould be impaired by the repeated occurrence of ground motions at the OBE level. The plant should beshut down safely and brought back on-line when it is shown to be safe to do so. The US Regulationsuse a 3x103yt event to define an OBE. This is more restrictive at face value but comparison should bedrawn against the UK equivalent.

I.D. Discharges

BNFL plants are designed to minimize effluents (aerial/liquid/solid) in accordance with the ALARPprinciple. Additionally the design must be demonstrated to meet the plant allocation of the overall siteaerial and liquid effluent discharge authorizations.

I.E. Summary'

The BNFL system of plant design differs from the US approach of design-evaluate-fix in that safetyassurance is demonstrated against all stages of the design by use of the appropriate standards and safetyevaluation at conception, definition, build, commissioning and operate stages. The final design isvalidated against the Company criteria.

The additional BNFL criteria particularly relating to low consequence events may incrPxtsethe extent ofsafety protection systems required for the plant.

4.1.3-3

I.F. Cost Effects

In general the equipment requirement associated with safety and environmental protection would besimilar for both the US and UK locations. Certain differences may occur in the engineering of therequirementswhich could lead to cost differences. These areas are listed below.

Control System. Within the UK the regulatory bodies will not allow credit .to betaken for the use ofProgrammableElectronic Systems (PES) in the design of protectivesafety systems. This has resultedina dedicated hardwiredprotectivecontrol system being incorporatedinto SMP which is additionalto thenormaloperationcontrol system. This hardwiredsystemcomprises approximately250 circuitswhich inthe mainare duplicatedwithinthe normal operationalcontrolsystem. If the US regulatorswere to allowthe use of PES in the design of protective systems then cost reductions could be made in this area. Inadditioncertain BNFL criteria relating to lower consequenceevents necessitates the need for additionalprotective equipment. As these are not requiredby the US regulations cost savings could also be madehere also.

Extreme Weather/Seismic. The design criteria for extreme weather and seismic events are based oninformation relatingto the areasurroundingthe SellafieldSite. These may differ significantlyfrom areaswithin the US (either more or less severe). As such cost effects could also vary as these criteriaare usedin the design of the overall building structureand also the internalplant and equipment. In the absenceof any details on plant location within the US no furtherinformationcan be provided.

4.1.3-4

4.1.3-2 , SAFEGUARDS

BNFL has worked with the safeguards regulatory authorities (IAEA and Euratom) for many years.forming a good workingrelationshipanda full understandingof the objectives and requirementsof eachorganization. BNFL's experience in safeguardsdates back to the mid-seventies and the Company hasbeen a leading participantin major internationalsafeguardsprojectsandvariousspecialized advisory andconsultants' groups [e.g. SAGSI (Standing Advisory Group on SafeguardsImplementation),LASCAR(Large Scale Reprocessingproject)and is representedon the Steering Groupof ESARDA (the EuropeanSafeguards Researchand Development Association)].

The Company has since 1977. operated a policy of including safeguards in the design criteria of newplants. BNFL has many years experience of designing plants to take account of these safeguards andmaterials accountancy obligations. This experience has been brought to bear on the design provisionsfor materials a_':countancyand safeguards provisions in the Sellafield MOX plant.

SMP has been designed to satisfy the IAEA criteria which state that it must be possible to detect aprotracted loss or gain of 8 kg plutonium over oae years operation (in process areas) or an abrupt lossof 8 kg plutonium within one month. These criteria apply to bulk materials. In the case of discrete items.the inspectorates must be able to detect any item loss or gain.

The safeguardsapproach for SMP has been designed to be of the highest standardand to meet all national(UK) and international (Euratom/IAEA) requirements. SMP may be designated for inspection by theIAEA. The IAEA approach to automated MOX fabrication plants favors the use of Near Real TimeMaterials Accountancy(NRTMA), in-line Non-DestructiveAssay (NDA), and advanced containmentandsurveillance employing surveillance, safeguardsseals, and weighing. The currentdesign provisions thatare being incorporatedinto SMP will ensure that the plant satisfies all the requirementsfor internationalsafeguards.

Cost Effects. Due to the close similarity of the US and UK requirements for safeguards no costimplications are envisaged in this area.

4.1.3-5

4.1.3-3 SECURITY

In the UK minimum standards are laid down for the physical protection of nuclear material which fullymeet the recommendations of the InternationalAtomic Energy Agency (IAEA). These are published asINFCIRC/225/Rev.3 under the title "The Physical Protection of Nuclear Material" and the mandatoryrequirementsof the "Convention on the Physical Protection of Nuclear Material" (INFCIRC/274/Rev. 1)which the United Kingdom has ratified.

In accordance with IAEA recommendations, nuclear materials are categorized according to theirsensitivity which depends upon the type and quantity of material see Table re.L3-1 The category intowhich material is placed is the criterion for the minimum standardof physical protection it is accorded.

As SMP contains Category 1 quantities of nuclear material the physical security provisions are designedaccordingly. The SMP security provisions are detailed within a Security specification. This has beendeveloped in conjunction with the BNFL security advisors to ensure the design of SMP complies withthe appropriate regulations. This document is of a confidential nature and cannot be reproduced here.However, listed below are the type of provisions provided for plants at the Sellafield site. (Note SMPfalls into the inner area criteria_.

Overall, protection is achieved at sites which contain nuclear material by a perimeter fence. This ispatrolled by the United Kingdom Atomic Energy Authority Constabulary (UKAEAC), with access to thesite controlled by a pass system for personnel and a gate or barrier arrangement for vehicles.

To implement a "defense in depth" principle and to concentrate defensive measures where they are mosteffective, three types of areas have been defined for sites holding nuclear material. Progressively throughthese three areas more stringent steps are taken in terms of physical protection measures and in therestriction of access to those who really require it and whose trustworthiness has been established. Thethree types of protected areas may be defined as:

a) Inner Area - created where Category I quantities of nuclear material as defined in INFCIRC225Rev.2 are used or stored and where therefore the consequences of theft or dispersal of the materialare such that extremely rigorous restrictions are required.

b) Intermediate Area - created where Category II and some Category III quantities of material are usedor stored and where although the consequences of theft or likelihood of dispersal of the material arenot so serious as for Category I quantities, rigorous restrictions are nevertheless required.

c) Outer Area - all other parts of the site not specifically identified as Inner or Intermediate area.Protected by the site perimeter and control of entry system.

In addition, Vital Areas are identified and established. These are areas identified by safety specialists ascontaining equipment systems or devices which are, alone or in combination, vul,erable to sabotage, theeffects of which would be sufficient to cause a radiological hazard to the public. Access to Vital Areasis limited and controlled.

The prime physical protection features of Inner Area protection include:

a) a controlled access point with a unique identification system.

4.1.3-6

b) a physical barrier around the building, patrols by armed police within radio communication to thecentral control post, intruder detection systems both external and internal, CCTV, and effectivelighting.

c) nuclear material monitoring equipment.

d) specific standards of construction for stores or rooms where material is held.

The prime physical protection features of Intermediate Areas include:

a) a fence or structure of a building delineating the area so that entry and exit are effectivelycontrolled.

b) a checking system at entry points to allow entry only to authorized persons.c) the provision of an intruder detection system and appropriate area lighting, supported by police

patrols and response arrangements.

Using the above measures as minimum standards the physical protection system for each site is designedspecifically for that site taking into account such factors as facility design, geographical location, and theform of the material being handled.

Many of the safety features incorporated in the design of plants, make sabotage difficult. The nature of

plants and processes adopted in the nuclear fuel cycle mean that most sensitive nuclear materials are ina form that makes them highly unattractive and the fact that they need to be heavily contained alsominimizes the consequences of sabotage.

Comparison of the above with US requirements is difficult due to the confidential nature of the detailedinformation. As the US and the UK standards are both based on the IAEA regulations the view is that

the provisions currently incorporated within SMP would be similar to those required by the NRC if theplant were to be constructed within the U.S.

On this basis it is felt that the security provisions would have minimal effect on cost and program if aplant similar to SMP were to be constructed within the US.

Cost Effects. Due to the close similarity of the US and UK regulations on security no cost implicationsare envisaged in this area.

4.1.3-7

TABLE 4.3.1-1

CATEGORIZATION OF NUCLEAR MATERIAL

.i

I 11 Ill(c)ii m

1. Plutonium (a) Unirradiated(b) 2kg or more Less than 2kg but 500g or less butmore than 500g more than 15g

2. Uranium-235 Unirradiated(b)

uranium enriched to 5 kg or more Less than 5kg but lkg or less but120%235U or more more than lkg more than 15kg

uranium enriched to 10kg or more Less than 10kg10% 235U but less than but more than20% lkg

uranium enriched above 10kg or morenatural, but less than10% 235U

i i ii

3. Uranium-233 Unirradiated (b) 2kg or more Less than 2kg but 500g or less butmore than 500g more than 15g

"41Irradiated Depleted ornatural uranium,thorium, or low-enriched fuel(less than 10%fissile content)

!(d)(e), ,,

(a) All plutonium except that with isotopic concentration exceeding 80% in plutonium-238.

(b) Material not irradiated in a reactor or material irradiated in a reactor but with a radiation level equalto or less than 100 rads/hour at one meter unshielded.

(c) Quantities not falling in Category III and natural uranium should be protected in accordance withprudent management practice.

(d) Although this level of protection is recommended, it would be open to States, upon evaluation ofthe specific circumstances, to assign a different category of physical protection.

(e) Other fuel which by virtue of its original fissile material content is classified as Category I and IIbefore irradiation may be reduced one category level while the radiation level from the fuel exceeds100 rads/hour at one meter unshielded.

it,

4.1.3-8

4.1.3-4 REGULATORY REQUIREMENTS

This section outlines the regulatotVprocesses requiredfor the constructionof nuclear facilities in the UKand the US.

Table 4.4-1 lists areas of regulationand the agencies within the US and UK who have responsibilities ofregulation in each area. All UK regulatory authoritiesare statutory bodies just as in the US. None are

private concerns and they regulate both civilian and other government agencies. In the followingdiscussion it is assumed that the readeris familiarwith US regulatory institutions and practices.

Just as in the US, the British regulatorsare broadly split into those concerned with safety and thoseconcerned with impact on the environment.

I. SAFETY

I.A. Nuclear Installations lnspectorate (Nil)

For safety, the Nil, in an exact parallel to the US NRC, has jurisdiction over all nuclear safety issuesconcerned with the operation of a nuclear facility from initial design and site selection todecommissioning. It is the Nil which grants and monitors compliance with the site licenses for allnuclear sites.

The responsibilities of the Nil. as incorporatedin a site license, include, but are not limited to:

Arrangements for emergency situations (eg. a majorrelease of radioactivity)Consignment of nuclear matter (no materialcan be c¢,nsignedto any place other than a relevant sitewithout Nil permission)Recording and investigating site incidentsAppointment of Duly Authorized PersonsNuclear Safety CommitteesPlants - plans, designs and specificationsOperating rules and Operational Safety AssessmentsSafety mechanismsPeriodic shutdowns

Decontamination and decommissioning

Before any major tasks can be undertaken it is necessary to involve the Inspectorate and no work onfacilities can commence without written authority.

The Nil has the authority to close down the operations of a nuclear site or facility should the operatorbe 'out-of-compliance', or to prevent the restart of a facility after shut-down or modification if they arenot satisfied with the work undertaken either in quality or completeness. Furthermore, the Nil has theauthority to conduct regular inspections or complete audits of site operations for compliance with licenserequirements. In all these respects the Nil is entirely similar to the US NRC.

However, in the US, since the US NRC has no jurisdiction over the US DOE nuclear facilities, or USDOD facilities, the regulatory picture is different for these facilities and contaminated sites. Thisdifference of safety standards between civilian and government sites does not exist in the UK.

4.1.3-9

I.B. Her Majesty's lnspectorate of Factories

While the Nil is responsible for nuclear safety, compliance with conventional industrial safety legalregulations rests with Her Majesty's lnspectorate of Factories although in certain instances the Nil willtake on the conventional safety role on behalf of the Inspectorate as a matterof convenience.

Her Majesty's Inspectorate is responsible for industrialcompliance for safetywithconventional industrialequipment in a directly equivalent manner to the Occupational Safety and Health Administration (OSHA).Likewise the Inspectorate has the responsibility for monitoring operator compliance in the routine testingof safety equipment and for the investigation of major safety incidents, industrial inquiries, etc.

Like the Nil, Her Majesty's lnspectorateof Factories has powers of enforcementincludingthe authorityto close down operations, or place prohibition notices on plants or equipment if they believe there is athreat to safety. This extends to prevention of equipment start-up if the agency is not satisfied with thework undertaken.

II. ENVIRONMENTAL IMPACT

II.A. Her Majesty's Inspectorate of Pollution (ItMIP)Ministry of Agriculture, Fisheries and Food (MAFF)

One significant difference between the US and British regulatory, process is that in the US differentagencies (the NRC and the EPA) are responsible, respectively, for civilian radioactive, and for non-radioactive (Hazardous) emission and wastes, whereas in Britain the same two regulatory agencies (HMIPand MAFF) working together have responsibility for both typgs of material. In both countries localauthorities also have an involvement.

Despite its overall federal responsibility for radioaaive emissions, the US NRC has no jurisdiction overthe US DOE nuclear facilities, so that the environmental compliance picture is different for these facilitiesand contaminated sites from similar civilian nuclear facilities, just as it is different for safety issues. TheEPA now has provided an element of commonality by providing environmental emission standards forUS DOE compliance and at the same time assisting the US NRC in its emission standards. Furthermore,in the US, the States have delegated responsibility for certain pollution controls and regulations. Thiscompetition between agencies does not exist in the UK.

Radioactive discharges are controlled, in the UK, under legislation administered by Her Majesty'sInspectorate of Pollution (HMIP) and the Ministry of Agriculture, Fisheries, and Food (MAFF). Theseregulators are jointly to limit gaseous or liquid discharges of both nuclear and non-nuclear hazardousmaterials through implementation of discharge authorizations which specify maximum authorized limitsand other conditions and requirements to be met by the operator. For example, this could be theestablishment of a minimum specified environmental monitoring program to assess the impact ofdischarges on the environment or on 'critical groups'. The limits, conditions and requirements arereviewed and reauthorized on a regular basis of one to three years. Breaches of the authorizations arean offense under the law and can result in prosecutions and substantial fines imposed on the operator.

III. USE AND DEVELOPMENT OF LAND

In addition to regulating nuclear and conventional safety, and the impact of operations on theenvironment, the UK also regulates the use and development of land. This is achieved through the

4.1.3-10

applicationofplanninglawsfortheconstructionofnuclearand non-nuclearfacilitiesina complexprocess.

Initially, this control is exercised through the Local Authority. When construction work is major, e.g.,for a nuclear facility, the planning approval sought of a Local Authority is generally referred directly tothe Department of the Environment who will either decide on the application or refer it to a 'PublicEnquiry'. The Public Enquiry allows the case for, and the objections against, the trajor constructionactivity to be heard. Such enquiries are chaired by a Government appointed inspector, who makes thefinal recommendation to the Secretary of State. In a number of cases such enquiries have lasted up totwo years (eg. for a nuclear power plant at Sizewell). The Public Enquiry process has its US equivalentin the Licensing Board hearings during the construction permit process for nuclear plants.

IV. STEPS IN THE REGULATORY PROCESS

Within the US the process of building a nuclear facility is different if the facility is a civilian one subjectto US NRC regulations than if it is a US DOE facility subject to DOE Orders.

On the civilian side the process requires three principal steps:

Approval of a site supported by an Environmental Assessment (approval being signified by thepublication of a US NRC Environmental Impact Statement)Approval of construction supported by a Preliminary Safety Analysis Report by NRC

. Approval of operation supported by a Final Safety Analysis Report by NRC

In practice the latter approval steps may be granted only in part as the NRC regulators become assuredof the safety of the facility and the hurdles of public hearing are overcome.

On the US DOE side the process is similar but different since DOE itself regulates its own safe operation.The process also requires three steps:

Based on a selected site DOE provides an Environmental Impact Statement for the facility whichin recent years has to comply with EPA regulatory emission control standards.

Construction is supported by a Preliminary Safety Analysis Report approved by a different sectionof US DOE, and

Operation supported by a Final Safety Analysis Report again supported by a regulatory section ofthe US DOE.

Because some nuclear facilities are of relatively low hazard (especially those handling low andintermediate waste) a Hazard Classification is performed at the PSAR state tOassist in a graded approachin which some facilities would be subject to full scrutiny and others would not. In a crude sense this haselements of a risk approach to regulation. The US NRC is not involved in these DOE approval steps.

Within the UK the process of building a nuclear facility is subject to one more formal step than in theUS. The steps are:

Approval of land-use through application to the local authority which would be referred to theDepartment of the Environment. This is supported by an Environmental Report containing anassessment of potential hazards to both the population and the environment, as well as the aesthetics

4.1.3-11

of the building, effect on local economy etc. equivalent to the Environmental Impact Assessment,It is at this stage that a possible Public Enquirs,'may be held.

Approval of construction supported by a Precommencement SafeD' Report (PCSR) by the NuclearInstallations lnspectorate (Nil). Risk-based methods are employed from the conceptual designonwards.

Approval of inactive and active commissioning supported by the Pre-Commissioning Safety Reportby Nil.

Approval of operation supported b.vthe Pre-OperationalSafeD' Repor_or the Plant Safety Case(PSC), equivalent to the FSAR. by Nil.

The third of these steps is additional to the formal civilian process in the US although in practice itgenerally occurs as final approval may be granted in stages through commissioning and low powertesting. Since Nil is an independent regulator)' authorit,v the formal process of regulation for newgovernment nuclear facilities is ver), parallel but a little stricter in the UK.

V. CONCLUSION

Whilst the regulator)' bodies which exist in the US and UK are different the overall requirements aresimilar. As such no cost or schedule implications are envisaged due to the regulator approval processes.

4.1.3-12

TABLE 4.4-1

REGULATORY BODIES IN THE US AND UK

Area of Regulation .... [ US Regulatory Authority [.... UK Regulatory AuthoritySafety ' ' ' .....

Licensing of nuclear sites' and " 'Nuclear Regulatory Commission Nuclear installationsfacilities (NRC) Inspectorate (Nil)

Agreement State Authorities Her Majesty's Inspectorate ofFactories

. ii

Environmental Impact

Control of emissions and Nuclear Regulatory Commission Her Majesty's Inspectorate ofreleases of radioactive pollutants (NRC) Pollution (HMIP)

Agreement State Authorities Ministry of Agriculture, "Fisheries, and Food (MAFF)

National Rivers Authority

Control of emissions and Environmental Protection Her Majesty's Inspectorate ofreleases of non-radioactive Agency (EPA) Pollution (HMIP)(hazardous) pollutants

Ministry of Agriculture,Fisheries, and Food (MAFF)

State Authorities Local Authorities

National Rivers Authority

Use of Land

Approval of construction and Nuclear Regulatory Commission Department of the Environmentclosures of nuclear and non- (NRC) for nuclear facilitiesnuclear sites

Local and State Authorities forothers

4.1.3-13

4.1.3-5 SMP OUTLINE PROJECT SCHEDULE

i : i

1992 _ 1993 _ 1994 1995 i 1996 1997- .......................................... _ ..........

Q3 Q41Q1 t212 I213 Q41Q1 Q2 Q3 Q4 Q1 Q2 Q3 Q41Q1 Q2 Q3 I_}Q1 Q2 Q3 Q4i :

........ • .......................... -4 -..4-- ...........

OPERA_

: (SEE NO'I1E)ADVANCED

FUNDSMAIN B&C"CONST _ _TIONV V"P Vc=,,_._ V_

AGREE TO COMMENCE HANIX)VER FORCONSTRUCTION V V AGREE TOCOMMENCE _7.E_..._,.[_.,,o. co=,.ss_,_.

i 15,StlEPCBR ISSUEI_C

"_ E.qIIMATETO\/ _l \/ _7 ISSIJEI:ISC.41,_IVEV Vm AND APPROVAl S

i,,,,.=J_

[ IF51(;NMECH i PRFOR MAJOREQUIP

CONTRACTS: CONTROLSYS DELIVERAB.ES

_,._,_.V V )V V_,_,_ V V ,.,_=,,--_,._,_PROCUREMENT .......... _. i

CCUaENOE.et.t[_,=:;_ _ m_ i :CONSTRUCllON STRUCI'UREV VtlEAlrHEIIlrlGNT ) :"

V MElUINST/I .LkTION

IIISTAI [ A'IION ........................

_ FCmACTWEt-7PREHANDOVER TESTING _ VAND COMMISSIONING

........................

N.Io SIHne hmlassembles _1 Im available I_iOeIo Ihisdale airing from Ildive colnmissiomingaclktili_.

ii

i .

4.1.3-6 INFRASTRUC_ REQUIREMENTS

SMP is located within BNFL's Sellafield site at West Cumbria. As such it utilizes many of the facilitiesavailable within the adjacentTHORP complex and on the Sellafield site generally. Listed below are themain areas where existing infrastructure/facilitiesare utilized.

!. USE OF EXISTING BUILDINGS/STRUCTURES

Due to the close proximity of SMP and the THORP complex, it has been possible to utilize certain areasof THORP to house SMP equipment. These areas include:-

a) THORP West Annex - This area is approximately1600m2and is used to house part of the Heatingand Ventilation equipment, namely the CI supply and extract fans and the C2 extract fans.

b) THORP STACK - The THORP stack is being used to house the SMP C3 and C5 discharge flues.c) General Storage Areas - Several of the general store areas within THORP will be used by SMP

as a location for equipment spares, consumablesetc.

II. USE OF EXISTING PIPED SERVICES

The following services are available to SMP from the existing site facilities:

• LP Steam• Compressed Air• Breathingair• Argon• Nitrogen• Hydrogen• Demineralized Water• Domestic Water• Condensate Disposal• C2/C3 Effluent Disposal• Electricity

III. USE OF SITE FACILITIES

The following facilities are used by SMP

a) THORP Central Control Room - The THORP control room is used to control and monitor theSMP plant environmental systems (eg. activity in air. ventilation etc). In addition this area will actas the incident control center for SMP.

b) Maintenance Facilities - SMP utilizes the workshops and decontamination/maintenance facilitiesavailable within the THORP complex.

c> Sample .Mmlysis - SMP utilizes the existing site laboratories to analyze powder and pellet samples.The samples are automatically transferred via the existing THORP sample transfer system.

d> Chaagerooms/Access Control - SMP personnel enter via the THORP complex. As such theyutilize the existing THORP changerooms and security control systems.

4.1.3-15

e) Waste Treatment Facilities - SMP utilizes the comprehensive waste treatment facilities availablewithinthe Sellafield site. These facilities cover PCM, intermediate, and low.level waste packagingand storage.

The following general site facilities are used by SMP:

a) Site Police Forceb) Site Fire Brigadec) Site MedicalCenter

L

4.1.3-16

4.1.3- 7" DEVELOPMENT REQUIREMENTS

The SMP process for manufacture of MOX fuel is based on that used within the existing MOXDemonstrationFacility (MDF). As such development of the basic process is now complete. However.further development work is underway to optimize SMP design and this can be split into 2 main areas:

• Development work to supportplant performance.• Developmentwork to supportequipment design.

1. PLANT PERFORMANCE

Although developmentwork to prove the main process is complete " ' + • " ._program is nowin place to provide plantoperatingdata. This work will involve varyir.. _'. _,:, s_parameters in orderto optimize the plantperformanceand establish acceptable tolerances f_c proc_uctquality.

II. ENGINEERING DESIGN DEVELOPMENT

This work will take the form of detail studies to prove engineering concepts prior to detaildesign/manufactureof equipment. Typical areas were this type of work is being carriedout are:

• Special radiometric instrumentation• Rod and Pellet inspection• Powder transfer• Remote maintenance

4.1.3-17

4.2 ADAPTING COMMERCIAL MOX FUEL

FABRICATION EXPERIENCE

A Hazard and Operability Study (HAZOP I)_ study was held to discuss the implications of

processing plutonia derived from weapons or 'A' grade plutonium using a commercial MOXfabricationprocess such as that currentlybeing developed for the Sellafield MOX Plant (SMP).

The SMP process is based on the conversion of civil plutonia powder (arising from the

reprocessingof fuel from MAGNOX and Thermal Oxide Nuclear.Power Stations) toMixed

Oxide (MOX) Fuel elements. Figure4.1.2.-1 indicates the main features of the proposed SMP

processand this was takenas the processbasis for the study.

HAZOP is a rigorous systematic process used principally to identify potential hazards and

assess the safety of plant designs.

4.2-1

4.2.1 Objectives of Study

The objectives of the meeting wereto:

a) Identify the key assumptionswhich neededto be madefor thestudy. The assumptions

relatedto the form of 'A'grade plutonium,its isotopic compositionetc.

b) Identify the key issues (in terms of safety, operabilityand process)which would need to

•be addressedin adaptinga commercialMOX fuel fabricationprocess to handle 'A' -

gradeplutonia.

4.2-2

4.2.2 Assumptions

The following assumptions have been made with regardto the form and isotopic composition of

the 'A'gradeplutonium:

a) 'A'grade plutonium contains 5 w/o Pu-240. The sensitivity of the study to plutonia,with Pu-240 content of < 5 w/o would also be considered.

b) The process Willreceive plutoniapowder. The conversion of weapons plutonium.to the

oxide form will occur in a facility providedelsewhere.

c) The moisturecontent of the plutoniapowderis 1.5 w/o.

d) The plutonia powder is 5 years aged. If the age is increased the dose rates would

increase due to Am in growth and the heatoutputwould increaseslightly.

e) The HAZOP I study is based on the assumption that civil grade plutonia is replacedin

terms of throughputwith plutonia,derivedfrom 'A'gradeplutonium

A comparison of these assumptions with the current SMP Reference Case is given in Table4.2.2-1.

Table 4.2.2-1. Comparisonof SMP Reference Fuel with Assumed 'A'GradeFuel

SMPREFERENCE ASSUMED'A' GRADEPuO,. PuO:

Bum Up 45 Gwd/teU 500 Mwd/teU

Aging 5 years 5 years(plus5 year cooling priortoreprocessing)

Pu-240 Content 10 W/o 5 W/oi

Moisture 1.5 W/o 1.5 W/o

4.2.3 HAZOP Study

The HAZOP I study was undertaken using the key word listing given in Table 4.2.3-1 The basis

of the study was the proposed SMP process flow diagram shown m Figure 4.1.2-1.

Table 4.2.3-2 is a record of the discussions from the meeting for each of the key words listed inTable 4.2.3-1.

A single action was generated against Nuclear Technology to determine the isotopic composition

of typical 'A' grade plutonium. This action (Action 1.1) and its response are included in theHAZOP minutes in Table 4.2.3-3.

4.2-t

Table 4.2.3-1. Key Word List for the HAZOP I

EXTERNAL DOSE SHIELDINGINTERNAL DOSE LOSS OF CONTAINMENTVENTILATION EFFLUENTSCRITICALITY FIREEXPLOSION/DETONATION IMPACT DAMAGEMIXING OF FEEDS SEISMICEXTREME WEATHER LOSS OF SERVICESINSTRUMENTATION/INTERLOCKS MAINTAINABILITYTOXICITY CORROSIONDECOMMISSIONING DOMINOHUMAN FACTORS OTHERS

4.2-5

Table 4.2.3-2. HAZOP I Record Sheet

Meeting No: 1Date: 6/12/93

Keyword Discussion Action/Recommendation

ExternalDose/ The anticipateddose rates from the 'A' grade PuO2 Action 1.1 - On D.Shielding will be lower than those being emitted from PuO: Winstanley (Nuclear

derived from civil reactors. The response to HAZOP Technology. Determineaction 1.1 (see attached)provides a comparison with the isotopic composition ofthe expected gamma and neutrondose rates between weapons ('A') gradecivil and A grade PuO2. It is seen for A grade PuO, plutoniumthat will bedose rates are lower, handledwithin the

proposed MOX productionConsequently this reduction in external dose rates facility.may enable the MOX fabrication in the proposedplant to involve more manual intervention.Additionally. the required shielding may be able to be

reduced.|

Internal It is anticipated that 'A' grad PuO: will be lessDose/Loss of radiotoxic than that used in the SMP process;.Containment However. the PuO, handled will never the less be

extremely radiotoxic and similar precautions willneed to be undertaken to ensure that its primarycontainment is maintained at all times.

Ventilation The requirements as indicated above are for all PuO2bearing material to be maintained in primaE/containment. In some instances, containment wouldbe provided by the various ventilation systems (e.g.,C5 extraction on the glovebox system and the C3/C2for secondary containment. These ventilation systemswill need to be provided even though theradiotoxicity of 'A' grade PuO: is less than that forcivil grade PuO:.

The decay heat load of 'A' grade PuO, will be lessthan that from civil grade PuO:. This may reduce theneed for additional cooling or the requirement forcooling to be provided by the ventilation sy,,;tem.

Effluents No additional liquid/solid effluents due to h;mdling of'A' grade PuO2 rather than civil grade PuO: wasidentified.

4.2-6

Table 4.2.3-2. HAZOP I Record Sheet (Continued)

Meeting No: 1Date: 6/12/93

- i....

Keyword Discussion Action/Recommendationi i |li i i ii iHi

Criticality It was noted by Nuclear Technology that thereduction of the Pu240 content of the 'A' gradematerial would result in an increase of the reactivityof the vessels. Nuclear Technology indicated that areduction of the Pu240 content from 10w/o to 5w/oreduces the safe mass bv 5 % and it is concluded thatto maintain the same level of safety, the proposedprocess vessels should be reduced by a similarfraction, though the actual reductio needs to beconfirmed.

This may have an effect on the plant capacity, thoughthe actual magnitude is unclear. However, should thecapacity be reduced, then this may be offset by howthe plant is operated, controlled, or by additionallines.

Fire/ As previousiy indicated (under ventilation) the heat....Overheating generation capacity of 'A' grade PuO2 is less than

that for civil grade PuO,. and consequently theoverheating hazard potential should be reduced.

iii i ii

Explosion/ No additional explosion or detonation hazards wereDetonation identified due to the processing of 'A' grade PuO,

rather than civil grade PuO:., ,1

Impact Damage No additional impact scenarios were identified due tothe processing of 'A' grade PuO_ rather than civilPuO:.

Mixing of Feeds The potential option that th'e plant may also processcivilian PuO: was identified. Should this occur thenincreased dose rates compared with 'A' grade PuO2can be expected. Consequently, should civilian gradePuO: be handled in addition to 'A' grade PuO: theneither additional shielding and/or lower celloccupancies would be required to maintain wholebody dose levels to acceptable levels.

Seismic No additional seismic requirements have beenidentified due to the processing of 'A' gradeplutonium rather than civil plutonium. The existingprovisions would need to be maintained.

Extreme Weather No additional design requirements have beenidentified due to the processing of 'A' gradeplutonium rather than civil plutonium. The existingprovisions would need to be maintained.

4.2-7

Table 4.2.3-2. HAZOP I Record Sheet (Continued)

Meeting No: 1Date: 6/12/93

Keyword Discussion Action/Recommendation

Loss of Services iThe lower decay heat from 'A' grade PuO,.rather .......than civil grade PuO: may reduce the need to provide!cooling following a loss of services.

!instrumentation/ The reductionin the Pu241 (and hence gamma rays) .....Interlocks may make it more diffict 't to detect PuO2in process

vessels. This may make the provision of safetyprotection systems, and product instrumentationmoredifficult to engineer.

Maintainability No additional'maintainability problems due to theprocessing of 'A' grade PuO2 rather than civil PuO:were identified. It was noted that the reduction in thedose rates from the fissile material should reducedose uptake from background contamination levelsduring maintenance.

Toxicity No additional"toxicity hazards due to the processing .........of 'A' grade PuO., rather than civil PuO2 wereidentified. Although radiotoxicity of 'A' grade PuO:is less than civil PuO, the containment requirementswould be the same.

i i

Corrosion No additional corrosion 'hazards due to the processingof 'A' grade PuO: rather than civil PuO: wereidentified.

Decommissioning No additional'decommissioning hazards due to the .........processing of 'A' grade PuO: rather than civilPuO: were identified. If the facility processed solely

'A' grade PuO: then dose rates duringdecommissioning would be less.

Domino It was noted that weapons grade material would not ....be in the form of PuO2 powder required for feed tocommercial MOX plants and that conversion to theoxide state (ie. PuO,.) will need to occur within anupstream plant. The HAZOP study does not considerthis conversion plant. However, it is noted that thecapacity of the conversion plant should match thecapacity of the MOX facility, i.e., the conversionplant should not impose a restriction on theoperability of the MOX plant.

Human Factors No additional human factor issues were raised as the "'

result of handling PuO., arising from 'A' gradeplutonium rather than civilian PuO,.

Others

Table 4.2.3-3. HAZOP Action Responses

ACTION: Determine the isotopic composition of weapons grade ('A' grade) PuO: that will be handledwithin the proposed MOX production facility.It is considered that the facility will handle PuO: that is approximately 5 years aged.

REPLY: The ratios shown below give a trend only and should not be used for specific calculationswithout verification that it is appropriate to do so.

'A' Grade Plutonium usually has a low Pu240 content (--5w/o).Nuclear Technology have in existence a MAGNOX FISPIN run for 500 Mwd/te irradiatedfuel. 5 yr aged. (PDEC 94) which has the following isotopic composition:

Pu240 4.3 w/o Pu236 9.52E-9 w/oPu241 O.19 w/o Pu238 3.8 E-3 w/oPu242 2.95E-3 w/o Pu239 95.5 w/o

A comparison between this composition and the SMP Shielding Design Basis (SDB) based onan arbitrary but appropriate system (i.e.. plutonium surrounded by 6ram steel) gives thefollowing ratios (A Grade/SDB) for 5 yr aged fuel.

RATIOS

3' n Total

Unshielded 0.05 0.11 0.05Shielded 0.11 0.11 0.11(6ram Steel)

The above ratios show that for these set of circumstances 'A' Grade plutonium give ---5 % ofthe unshielded dose rates and 10% of the shielded dose rates assumed from the reference case.

NB: Shielded dose rates are dominated by neutron dose rates. If the Pu240 content increasesthen the neutron dose rates will increase.

4.2-9

4.2.4 Main Findings

The following is a summaryof the mainissues raisedduringthe study:

a) Because the 'A'gradeplutonia contains fewergammaand neutronemitting isotopes (see

response to Action 1.1, Table4.2.3-3) the need for the fully automated processing could be

relaxed which may have particularbenefits for fuel assembly operations. This would not be

the case if the facilityhadto piocess bothcivil and 'A'gradeplutonium....

b) Though the radiotoxicityof'A' gradeplutonium is.marginallyless than civil plutoniumit

is consideredthat the containment integrity of any plutonium beating materialwill need to

be the same to that for a civil plutonia MOX plant.

c) The radiometricdecay heat arisingfrom plutonia derivedfrom 'A'grade plutoniumwillbe less than that from civilgrade material. Consequently this may reducethe requirementsto

providecooling of any plutoniumbeating materialholdup within the process.

d) The Pu-240 content of plutonia derived from 'A'grade plutoniumwill be less than the

corresponding Pu-240 content of civiliangrade material.This may require that the capacity

of any vessel thatmay hold pure plutoniapowder or MOX be reduced to ensure criticalitysafetyunder normaland faultconditions.Typicallyfor plutoniawitha Pu-240 contentof 10

w/o, a reductioninPu-240 contentto 5 w/o reducesthe safemass by S w/o.

The effects on the plant capacity because of this would need to be considered,though it is

....anticipatedthat any reductionin individualvessel capacitycould be off-set by for exampleincreasingthe numberof process lines etc.

e) It was noted that the lower Pu-241 content of 'A' grade plutonia will result in lower

external dose rates. As it is anticipatedthat some instrumentbased protection systems

within the plantmay, in some instances rely on the gamma rays being emitted from the

plutonia(eg. criticalityinstruments)thenthe availabilityof these systems could be affected.

f) The capacity of any upstream facility which may be used to convert weapons grade

plutonium to plutonia suitable for the commercial process should be such that it will not

restrict the operation of the MOX plant. In addition the upstream plutonium conversion

facility could incorporate a capability to blend the weapons grade plutonia with other

material (eg. Civil plutonia or urania) such that the criticality reactivity of the material was

no worse than civil plutonia. This would avoid increasing the difficulty in making a criticalitysafew case.

4.2-10

4.3 PLUTONIUM DISPOSITION COMPLEX INFRASTRUCTUREIN THE UNITED STATES

4.3.1 Introduction and Summary

Itwas concludedin Phase 1A that the infrastructureappearedto be establishedfor deploymentof

an ABWR Pu Disposition Complex in the United States. The brief surveysof each of the DOE

sites conducted underPhase 1Csupportthis conclusion.

TheDepartmentof Energy(DOE)has the sites andcapabilities,and theflexibilitywith these sites

and capabilities,to deploy anelectricpowerproducing,full MOX-fueledABWR PuDisposition

Complex. Forstudy comparisonpurposes,thereferencecase for deploymentof the Plutonium

Disposition Complex in the UnitedStates is a new "Greenfield," in which the facilities are

constructedand located all togetheron a hypotheticalsite at Kenosha,Wisconsin. Thepurposeof

the infrastructureportionof this studywas to determinethe extent to which the "complex"couldutilize the existing capabilitiesat one or moreof theexistingDOE andcommercialsites.

The studyincludedvisits anda collectionof dataforthe following sites:

• IdahoNationalEngineeringLaboratory(INEL)• Ne_,adaTest Site (NTS)• OakRidge Reservation (ORR)• PantexPlant• Savannah River Site (SRS)• HartfordSite• LawrenceLivermoreNational Laboratory (LLNL)• Los Alamos National Laboratory (LANL)

The study team confm'nedfrominterviewsand facility tours with DOE siteoffice and site

contractorrepresentativesthat sites, facilities, resourcesand capabilities alreadyexist at DOE and

commercialsites which can be used at a cost and schedule saving.

The fast five of these sites arealso being evaluatedfor operationalelementsby the nuclear

weapons complexreconfigurationprogram. This section covers these sites, and summarizesthe

capabilitiesfor conducting the Pu dispositionfunctions,and for managingthe interfacesbetween

the Pudisposition and the nuclearweaponscomplexreconfigurationfunctions,at each site. A brief

visit was made to each site. The introductionby GEfor each visit includedFigures 1-2plus a

descriptionof the ABWR plant. The discussionsthen coveredas many of the topics shownin

Figure2 in as muchdetailas allowedby the timeavailable.

4.3-1

In order to provide some consistency between the evaluations for the weapons complex and Pu

disposition programs, the degree of readiness at the site for each of the capabilities required for Pu

disposition was classified as being at one of following three levels, in a manner similar to that used

by the nuclear weapons complex reconfiguration program:

• EXISTING (E) - .Ca_.bility exists at the site and can be applied to Pu disposition withonly minor refurbishing required.

• UI_RADE (U) - Capability exists at the site and can be applied to Pu disposition withupgrading required, such as renovation of an existing facility and installing some newequipment.

• GREENFIELD (G) New capability required at the site.

The latter three sites, Hanford, LLNL, and LANL, although not considered operational site

candidates, have facilities or technical capabilities which could be supporting functions for Pu

disposition, and arc also covered in this section. The mission of the HartfordSite, although it

contains facilities applicable to Pu disposition, is believed to be primarily to implement

environmental restorations. LLNL and LANL were not considered for production-type activities

such as MOX fuel fabrication or reactor operation. Visits to these sites were made during and prior

to Phase IC. The results reported here include for completeness some of the conclusions discussed

in the earlier Phase IB report.

It is clear that considelable cost effective, installed capability is available within the DOE

community now for meeting the Pu disposition needs in the near term with an electric power

producing, full MOX-fueled ABWR plant. These capabilities can be implemented in the short term

(8-10 years) with effort ranging from minor refurbishing to upgrading of existing facilities, with

only a few requirements, such as the reactor, being Greenfield efforts at all sites. It is anticipated

that a minimum cost deployment will be to locate the entire Pu Disposition Complex at one site.

SRS, ORR, and INEL already have in place significant applicable elements. It is also possible to

take advantage of unique capabilities which exist at individual sites and create a distributed

"complex," with some additional cost for transportationbetween sites.

Section 4.3.2 provides a definition of the requirements for the Pu Disposition Complex,

and Sections 4.3.3-10 provide the results of the evaluations.

4.3-2

4.3.2 Pu Disposition Complex Site Requirements

This sectiondescribesthecapabilitiesandfacilitiesneeded bya site in orderto beconsideredfor

locationof some or allof theABWR Pudispositionfunctions,andthe capabilitieswhich the site

musthave for maintainingcriticalinterfaceswithotherprogran_(such asthe NuclearWeapons

ComplexReconfigurationProgram).

OVERALL SITE QUALIFICATION

A completelyGreenfieldsite for theABWRPu Disposition Complexis the baselinefor

comparativepurposes. Theactualsite selectedwill probablynot be a Greenfield,given that several

attractivesitesalreadyexist withinthe DOE infrastructure.Thecandidatesite mustbe sufficiently

largeand have thephysicalresourcesand technicalexperiencein nuclearprogramstoprovide the

space, utilities, and stafffor the operation,safety,security,safeguardsand maintenance of the Pu

disposition functions.

Pu FEED MATERIAL INTERFACE

The study was directed to assume that the weaponsprogramwill provide PuO2feed material for

Pu disposition in accordancewith the specificationsfor fabricationof MOX fuel for ABWR fuel

pins. The site must have the capability to receive this PuO2. A site with the capability for Pu

processing, and which therefore could also be the producer of the PuO2, presents the advantage of

minimizing the complexity and cost for PuO2 transportationfromthe interim storagesite to the Pu

Disposition Complex.

MOX FUEL FABRICATION

The Pu Disposition Complex includes a MOX fuel fabricationfacility whichreceivesand stores

PuO2and UO2, blends the PuO2and UO2, sintersthe MOX pellets, loads thefuel pins,

assembles the fuel assemblies, stores thecompletedassemblies, andpackages the assembliesfor

transportto thereactor. The site musthave the capabilityforPu processingwhich can be applied

to MOX fuel fabrication.A site with the capabilityfor both MOX fabricationandreactoroperation

presentsthe advantageof minimizingthecomplexity and cost for transportationof freshMOX fuelassemblies.

4.3-3

TRITIUM TARGET FABRICATION INTERFACE

It is assumed that the weapons program will provide the tritium targetrods for tritiumproduction in

accordance with the specifications for including the tritium targetrods in the MOX fuel assemblies

for the ABWR. The fabrication of the target components and the completed targetrods can be

accomplished by commercial vendors. The site must have the capability to receive the rods. A site

with the capability for integrating theprocurement of the components of the targetrods produced

by commercial vendors, and then completing fabrication of the rods onsite, presents the advantageof minimizing the complexity and cost of tritium target fabrication.

REACTOR PLANT SITING

The Pu Disposition Complex being proposed by GE includes a 1350 MWe ABWR which is

essentially identical to the ABWRs being built now in Japan. The candidate site must have the

capability for locating this reactor on or near the site. A site with the capability for locating both the

MOX fuel fabrication facility and the ABWR plant on or near the site presents the advantage of

mimmizing the complexity and cost of transportation of the fresh MOX fuel assemblies.

POWER TO THE GRID

The Pu Disposition Complex will sell the electric power generated by the ABWR plant to an

Independent Power Producer (IPP)/Utility customer. The candidate site must have the capability

for access to a commercial grid for transmission of the power offsite. A site in an area with a need

for electric power presents the advantage of maximizing revenue from sale of the power.

SPENT FUEL DISPOSAL INTERFACE

The study was directed to provide ten years of storage at the reactor site for spent fuel. This is

provided by the in-reactor spent fuel pool of the GE ABWR design. It is assumed that the spent

MOX fuel assemblies will be turned over to the U.S. Nuclear Waste Disposal Program after the

period in the in-reactor spent fuel pool. It is anticipated that temporary storage of the fuel

assemblies discharged from the in-reactor pool will be required pending direction for disposal.

The candidate site must have the capability for handling spent MOX fuel assemblies for transport to

this temporary storage location. A site with the capability for additional temporary spent fuel

storage on or near the site presents the advantage of minimizing the complexity and cost of

transportation of the spent MOX fuel assemblies to some other temporary storage location pending

f'mal disposal.

4.3-4

WASTE MANAGEMENT

The Pu Disposition Complex will utilize state-of-the-artmethods andprocedures to minimize the

amount of waste requiring disposal. The candidate site must have the capability for handling high

levei, transuranic, low level, hazardous, and mixed wastes, and for packaging the waste for

shipment to approved facilities for disposal, and for managing the interface with the US Nuclear

Waste Disposal Program. A site with temporary or permanent waste disposal facilities on or near

the site presents the advantage of minimizing the complexity and cost of transportation of thewastes.

TRITIUM RECOVERY INTERFACE

It is assumed that the irradiated tritium target rods will be turned over to the weapons program after

discharge from the reactor and disassembly from the MOX fuel assemblies. The candidate site

must have the capability to coordinate the transport of the rods to the tritium recovery facilities. A

site with the capability for tritium processing, which therefore could conduct the tritium recovery

operations, presents the advantage of minimizing the complexity of shipping irradiated tritium

target rods and transportation costs.

SAFEGUARDS AND SECURITY

The Pu Disposition Complex will require safeguards and security for the storage, handling,

processing, and transport of special nuclear materials (SNM) from the point of receipt of the PuO2

from the weapons program through insertion of the MOX fuel assemblies into the reactor core and

start of irradiation. After irradiation, the fuel assemblies will require safeguards and security

equivalent to that for spent fuel in commercial reactors. The candidate site must have the cat-ability

to provide for the accountability of the SNM and for protection against SNM diversion.

SAFETY AND ENVIRONMENTAl, APPROVAL

The Pu Disposition Complex will require safety approval by DOE and/or NRC, environmental

approval by the EPA, and federal, state, and local pemaitting agencies. The ABWR plant will soon

be certified by the NRC, and GE as the reactor manufacturer will carry the main burden of the final

safety approval process for the plant owner (i.e. either the government or an

IPP/utility). The MOX fuel fabrication plant is expected to be owned by DOE, with safety

approval by either DOE or NRC. The plant owner is responsible for obtaining the environmental

approval, and for obtaining the additional federal, state, and local permits required for operation of

the complex. The main support to the owner will be supplied by the reactor manufacturer and

4.3-5

architect/engineer, for the source terms and plant description, and by the site, for the environmental

conditions and the impact of operation of the complex on this environment. The candidate site must

have the capability to provide effective support to the approval process, including analyses,

documents, and participation in reviews as required. A web characterized and documented site

will present the advantage of cost effectiveness and low risk for safety and environmental

approval.

TRANSPORTATION

The Pu Disposition Complex will require access to the site and transport for SNM and wastes both

on-site and off-site. The candidate site must have the capability for receiving heavy equipment

during construction, for unloading and inspection of incoming SNM shipments, for transport of

SNM and wastes on-site between facilities and between protected areas, for packaging SNM and

wastes for off-site shipment, and for strict accountability of the SNM.

SUPPORTING SITE ASSETS

The Pu Disposition Complex will require support in addition to the areas noted above. The

candidate site must have a management organization experienced in large nuclear projects, a

technical and production oriented work force, adequate utilities such as water and power,

technology capabilities, strict quality assurance requirements, and a community attitude that is

favorable to Pu processing and reactor programs.

4.3.3 Capabilities at the Idaho National Engineering Laboratory

The Idaho National Engineering Laboratory (INEL) was visited on 15 December 1993. A full day

of discussions was held with DOE site office and site contractor management and staff to obtain an

overview of the capabilities at INEL applicable to a Pu Disposition Complex. The summary

presented in Table 1 is the result of these discussions, followup telephone calls, and review of

documents describing the INEL facilities. These results should be considered preliminary pending

thorough review with the INEL staff.

4.3.4 Capabilities at the Nevada Test Site

The Nevada Test Site (NTS) discussions were held with DOE management and staff at the DOE

Nevada Operations Office in Las Vegas on the morning of 8 December 1993. The summary of the

results is presented in Table 2, based on these discussions and review of documents describing the

4.3.6

NTS facilities. These results should be considered preliminary pending thorough review _,itt_ the

NTS staff.

4.3.5 Capabilities at the Oak Ridge Reservation

The Oak Ridge Reservation (ORR) was visited on 11 November 1993. The full day of discussions

with site contractor management and staff at the Y12 facility included a brief driving tour of the

Y12, K25, and X10 (ORNL) areas, the CRBR site, and the barge docking capability on the Clinch

River. The summary of the results is presented in Table 3, based on these discussions, foUowup

telephone calls, and review of documents describing the ORR facilities. These results should be

considered preliminary pending thorough review with the ORR staff.

4.3.6 Capabilities at the Pantex Plant

The Pantex Plant discussions were held with DOE management and staff at the DOE Albuquerque

Operations Office and the DOE Amarillo Area Office on the morning and afternoon, respectively,

of 9 December 1993. The summary of the results is presented in Table 4, based on

these discussions. These results should be considered preliminary pending thorough review with

the Pantex staff.

4.3.7 Capabilities at the Savannah River Plant

The Savannah River Site (SRS) was visited on 8-10 November 1993. The meeting was sponsored

by the DOE Savannah River Site Office, which also coordinated the participation of the DOE

Southeastern Power Administration (SEPA) in the meeting. Discussions were held with

management and staff of the site contractor, and representatives of SEPA and the Southern

Company (an IPP/Utility with interests throughout the southeast and beyond). The ,.,i_k included

several tours of potential facilities for Pu disposition. The results of these discussions, an earlier

visit (August 1993) devoted to MOX fuel fabrication capability, followup telephone calls, and

review of documents describing the SRS facilities, are presented in Table 5. These results should

be considered preliminary pending thorough review with the SRS staff.

4.3.8 Capabilities at the Hanford Site

The Hanford Site was initially visited during Phase 1B to evaluate the Fuels Materials and

Engineering Facility (FMEF) as a building which might accommodate the MOX fabrication

activities for the ABWR Pu Disposition Complex. The Secure Automated Fabrication (SAF)

4.3-7

facilities which were designed for the fabrication of Fast Flux Test Facility (FFTF) fuel are of

particular interest. Since the SAF line was designed for fast reactor fuel, the lower plutonium

concentration and larger pellets associated with an ABWR design preclude the use of most of file

equipment installed in the facility. Despite these different types of ceramic the process utilized in

the SAF line is similar to those being considered for the ABWR Pu Complex and the safety and

accountability procedure should be nearly identical. Some Pu handling technology is available

from weapons work completed in previous years which may be applicable to Pu disposition. The

information developed during site restoration activities will be directly applicable to the waste

handling associated with MOX fabrication and reactor operation. The plutonium processing

experience and capability at Hanford could provide technology support to the Pu Disposition

program, and the FMEF could be upgraded for ABWR MOX fuel fabrication.

4.3.9 Capabilities at the Lawrence Livermore National Laboratory

The Lawrence Livermore National Laboratory was initially visited during Phase 1B activities to

review the processing technology for converting plutonium metal to the oxide utilizing the

hydride/dehydride process which is utilized to remove plutonium from a retired weapon. The

activities completed by the laboratory have resulted in the design and demonstration of a workable

process. A great deal of effort has been devoted to the robotics systems necessary for remote

cjpcrationof the hydride/dehydride processes. Much of the technology will be applicable to remote

processing associated MOX fuel fabrication. Several visits were made to Lawrence Livermore

National Laboratory during Phase 1C activities to furtherreview plutonium conversion technology.

In addition, the processes being developed to handle the scrap and waste steams generated during

plutonium hydride/dehydride activities were reviewed. The Plutonium processing experience and

capabilities at LLNL could provide technology support to the Pu Disposition Program.

4.3.10 Capabilities at Los Aiamos National Laboratory

The Los Alamos National Laboratory (LANL) was initially visited during Phase 1B activities to

assess progress on the technology development activities for Pu-to-Pu02 conversion, and the

potential for fabrication of mixed oxide/gadolinia (MOX/GAD) test assemblies to confirm the

burnout rate of gadolinia in an ABWR. In addition to the burnout rate information, MOX/GAD

processing capabilities could be demonstrated and final fuel fabrication facility design information

developed. "lhe technology utilized at this site to handle plutonium containing scrap and waste

streams, and accountability and safeguards, were evaluated for applicability to an ABWR complex.

A brief review of the Automated Retirement and Integrated Extraction System (ARIES) conceptual

4.3-8

design information was provided because much of the technology to be utilized in the retirement of

plutonium weapons components is directly applicable to the ABWR Pu Disposition Complex. A

second visit was made on October 21, 1993, to review the storage requirements for plutonium

which had been removed from the weapons. The processing technology available for the

conversion of weapons plutonium to plutonium dioxide was discussed and necessary development

activities defined. LANL personnel have begun the development of a fiber optic technique for

analysis of plutonium content, plutonium isotopic, and impurities through the use of laser

technology. Since all of the electronics are located outside of the glove box areas, the technology

is particularly attractive for use in a MOX facility. Generally, these analysis techniques are orders

of magnitude more rapid than standard analytical chemistry techniques. The Plutonium processing

experience and capabilities at LANL could provide technology support to the Pu Disposition

program.

4.3-9

t

Plutonium Dispositon Study-Phase 1CI

DOESTATEMENTOFWORK

GENERALELECTRIC- Advanced Boiling Water Reactor

!

Task 4: Investigate deployment strategies.

Assess alternate deployment strategies Including useof existing DOE and oommeralal facilities, Integrationof fuel fabrication and reactor complex facilities,complex location, transportation and logisticrequirements for the plutonium feed material, wastestream products, and spent fuel.

FIGURE 1 GE STATEMENT OF WORK

J

@ THIS MEETINGII I

I ii

• Provide information on ABWR Plant and Progress of PlutoniumDisposition Study

• Obtain Information on existing Infrastructure for evaluation of feasibility,cost and schedule for deployment of Plutonium Disposition complex,inc,uding tritium production, and possible Integration with weapons

= reconfiguration programI

Pu Feed Material (_Jnterface) Waste ManagementMOX Fuel Fabrication Tritium Recovery (Interlace)Tritium Target Fab (InteHace) Safeguards and SecurityReactor Plant Siting Safety Approval

Power to the Grid (Intedace) Environmental Approval

Spent Fuel Disposal (Interface) Transportation

Supporting Site Assets

FIGURE 2 INFRASTRUCTURE EVALUATION OBJECTIVES

DRAFT 15Jan94 file IDEPLOY.

**************************** DRAFT TO BE REVIEWED WITH INEL ********************************

Table 1 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

G = Greenfield. New Capabilities Required.

U = Upgrade of Existing Capabilities Possible.

E = Existing or Planned Capabilities Meet Requirements.

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Overall Site Qualification E - Currently operating. Has served for "45 years as DOE site for part of

Nuclear Materials Production Complex, reactor R&D, and fuel processing.

INEL is still performing these missions on a reduced scale.

- Occupies "890 square miles -29 miles west of Idaho Falls, ID

- Eleven Technology Areas: Test Area North (TAN), Test Reactor Area (TRA),

Central Facilities Area (CFA), Radioactive Waste Nanagement Complex (RWMC)

Auxiliary Reactor Area (ARA), Power Burst Facility/Power Excursion

Reactor Test (PBF/SPERT), Idaho Chemical Processing Plant (ICPP), and

Argonne National Laboratory West (ANL-W), Idaho ReseaLch Center (IRC),

Idaho Supercomputer Center (ISC), INEL Engineering

_j PuO2 Feed Material Interface

' (IF w/Nuclear Weapons Complex Reconfig Program)

- Pu Receving and Storage E - Management and staff experienced with SNM handling including Pu

(within weapons program) - SNM vault in Fuel Processing Restoration (FPR) facility could receivePu metal or oxide

- Pu-to-PuO2 Conversion E - Management and staff experienced with Pu processing. Pu metal core

(within weapons program) fabricated and irradiated in Materials Test Reactor.

- Hot cells in FPR facility can be used for Pu-to-PuO2 conversion

- PuO2 Feed Interface Management E - Experience managing SNM is applicable to management of interface

with supplier of ?uO2 feed material whether or not Pu-to-PuO2conversion function is located at INEL

MOX Fuel Fabrication

- MOX Pellet and Pin Fabrication U - Experience with Pu processing is applicable to extension of capability

to MOX blending, sintering, and pin loading. [TBR]

- Fuel Processing Restoration (FPR) facility was completed in 1992, is

uncontaminated, and can be modified to house the NOX lines.

New equipment for powder blending, pellet sintering, and fuel

pin loading and closure is needed.

- FPR is six stories tall, 160,000 square feet floor space,

state-of-the-art radiation protection, contamination control,

decontamination, natural phenomena protection, remote handling,

50/5 ton bridge crane, DOE approved SNM vault

- ABWR MOX Assembly Fabrication U - FPR can be modified for MOX Assembly fabrication

- Fresh MOX Assembly Storage U - FPR can be modified for fresh MOX Assembly storage

Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Tritium Target Fabrication Inter_ace

(IF w/Nuclear Weapons Reconfig P_ ram

- ABWR Tritium Target Fabrication U -- Could be located in FPR together with or separate from MOX lab.

(within weapons program) New equipment for pin loading and closure needed.

- Tritium Target Fab Interface Management E - Experience in managing procurement from commercial vendors, and in

managing design and irradiation of NP-MHTGR tritium target development

program and irradiation of LWR tritium targets, is applicable

to management of interface with supplier of tritiue targets whether

or not ABWR tritium targets are fabricated on-site

Reactor Plant Site Qualification

- Reactor Site E - ABWR plant could be located on available open, unencumbered,

contamination-free 10,240 acre site (more if needed) completelywithin INEL boundaries. This site was selected for NPR.

- Commercial site could be made available outside protected zones

similar to WPPS plants on Hanford reservation

- Site Qualification E - Site characterization is completo at NPR site, where Pu Disposition

Complex would be located

- Reactor Plant G - ABWR plant will be a new facilityI

t_ - Reactor Plant Operation E - Management and staff experienced in reactor operation

- 52 special purpose reactors designed, construct*d and operated.

Currently 12 operable reactors at INEL.

Power to the Grid Interface

(IF w/IPP/Utility)

- Proximity to Commercial Grid G - New grid needed and in progress. Power transmission corridor from

mid-Idaho to Las Vegas area initiated by Idaho Power Company to handle

increase in power transmission from Canada, and originally expected

from NPR, currently in Environmental Impact Statement process with

Record of Decision expected in near future. Note that this corridor

will be completed regardless of any INEL activities.

- Reactor-to-Commercial Grid Transmission G - New transmission lines needed. Idaho Power will pr-vide transmission

services for power produced at INEL as part of electic power sales

agreement.

- Electric Power Sales U - Idaho Power proposed to lead formation of regional utility consortium

to construct and operate power conversion side of NPR and/or to

distribute electric power generated by NPR. Agreement on purchase

of power generated by Pu Disposition Complex by Idaho Power could

negotiated contingent upon pricing, availability, and reliability

of offered power.

- IPP/Utility Interface Management E - Experience managing large electric power distribution system for INEL

and past working relationship with Idaho Power is applicable to

managing interface with IPP/Utility for sale and transmission of

power generated by Pu Disposition Project

Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Spent Fuel Storage Interface

(IF w/Nuclear Waste Disposal Program)

- Spent Fuel Storage E - Environmental Impact Statement (EIS) in progress for storage of

(within waste disposal program) commercidl spent fuel (DOE Programmatic Spent Nuclear Fuel Management

and INEL Environmental Restoration and Waste Management Programs EIS).Schedule is draft Jun94.

- Currently available pools include Idaho Chemical Processing Plant (ICPP}and naval reactor facilities

- Construction of additional spent fuel storage facilities planned

- Manages development of storage casks. Conducted Dry Rod Consolidation

Technology (DRCT) project. Conducted testing and analysis of metal

casks for dry storage.

- Spent Fuel Transport E - Management and staff experienced in spent fuel transport, including LWR

(within waste disposal program) - Manages development of transportation casks. Cask Systems Development

Program supports acquisition of prototype casks for spent fuel

transport from reactors to repository.

- Manages transport of spent fuel, including LWR, to Hot Fuel Examination

Facility (HFEF) at INEL

- Manages transport of spent LWR fuel assemblies from reactors as part

of Spent Fuel Storage Cask Testing Program and other programs!

- Spent Fuel Interface Management E - Experience managing spent fuel operations is applicable to managementof interface with Nuclea_ Waste Disposal Program whether or not spent

fuel is stored or transported h_ INEL

Waste Management

- High Level Waste (HLW) E - Management and staff have recent experience in designing, constructing,

operating, and maintaining facilities to handle, process, and store HLW

- Waste treatment capabilities/facilities include handling, processing,

calcining, and storage

- Currently stored at Idaho Chemical Processing Plant (ICPP)

- New Waste Management Center is in planning process

- Transuranic Waste (TRU) E - Management and staff have recent experience in designing, constructing,

operating, and maintaining facilities to handle, process, and store TRU

- Retrievable contact handled TRU and TRU-mixed wastes currently

stored at Transuranic Storage Area (TSA) of Radioactive Waste

Management Center (RWNC)

- Retrievable remote handled TRU and TRU-mixed waste currently stored

at Intermediate-Level Transuranic Storage Facility (ILTSF) of RWNC

- Low Level Waste (LLW) E - Management and staff have recent experience in designing, constructing,

operating, and maintaining facilities to handle, process, and store LLW

- LLW disposal currently at Subsurface Disposal Area (SDA) of RWHC

Table 1 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Waste Management (ton't)

- Hazardous Waste E - Management and staff have recent experience in designing, constructing.

operating, and maintaining facilites to handle, process, and storehazardous wastes

- Mixed Waste E - Management and staff have recent experience in designing, constructing,

operating, and maintaining facilities to handle, process, and storemixed wastes

- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable to

management of _nterface with Nuclear Waste Disposal Program

Tritium Recovery Interface

(IF w/Nuclear Weapons Complex Reconfig Program}

- Extraction G - A new facility to extract tritium from targets irradiated in

(within weapons program} the ABWR Pu Disposition Complex is required

_ - Purification G - A new facility for purifying the tritium to specifications

_o (within weapons program) for shipment to the RTF at SRS is requiredI

- Reservoir Loading G - New facility required. It is believed the Replacement Tritium

(within weapons program) Facility at SRS has been designated by DOE for reservoir loading

function for the weapons program.

- Tritium Recovery Interface Management E - Experience managing tritium target development and testing at INEL

under NPR program is applicable to management of tritium recovery

interface whether or not some or all of the tritium recoveryfunctions are located at INEL

Safeguards & Security

- Accountability E - Management and staff experienced in implementing strict

accountability procedures for storage, handling, and transport

of SNM in all forms including uranium and plutonium Iota1,

oxide, components, and waste

- Protection U - INEL dedicated safeguards and security contractor experience, well

trained and equipped to provide necessary protection and initiate

effective emergency responsed to security infractions, is applicable

to providing additional personnel, fences, and guard posts for

Pu Disposition Complex

Table I (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Safety and Environmental Approval

- Safety Approval U - INEL has experience with safety approval by DOE of nuclear facilitiesincluding reactors.

- INEL technical staff provides assistance to Office of Nuclear Reactor

Regulation (NRR) in instrumentation and control systems, electrical and, xn sorvxco and pro-service inspectionmechanical components and systems " - "

amd testing of piping systems, equipment qualification, radiological

issues, operator licensing examinations and training programs, standard

technical specifications and plant specific technical specifications,

license renewal activities, ALWR issues, and thermal/hydraulic analysis.

Recent emphasis has been loss on operational aspects and more on

ALWR issues.

- DOE and NRC requirements are similar. Experience with DOE safety

approval procedures, and with technical assistance to NRR, is

applicable to extension of capability to accomodate NRC procedures.

- Environmental Approval E - Management and staff experienced with environmental approvalrequirements to obtain Record Of Decision (ROD) by DOE with EPA

concurrence for nuclear facilities including reactors.

- Site data developed for ROD issued to INEL for Environmental ImpactI

Statement for Special Isotope Separation Project, and other

C_ evaluations (EIS and EA reports listing available upon request),

is applicable to Pu Disposition Complex

- Federal/State/Local Permitting E - INEL has experience in permitting at all levels

Transportation- Site Access E - Access by air via Idaho Falls, ID

- Access by road includes two US highways and one State highway crossing

INEL boundaries. -230 miles on-site roadway classified principal arterial

and major collector routes. 500+ ton load capability on INEL roads.

- Access by rail is Union Pacific connection at Scoville Siding to

government-owned spur line linking developed areas within INEL. Gantrycrane at Scoville Siding and spur line handle up to 160 ton loads,

-15' high x 10' wide for heavy equipment delivery during construction.

- Unloading & Inspection E - Management and staff experienced in receiving SNM in all forms

on sxte transport in SNM- On-Site Transport E - Management and staff experienced in - "

in all forms within and between protected zones

- Two US highways and state roadway cross INEL boundaries

- Droposed ABWR plant and MOX fuel fabrication facility siteswithin -3 miles of each other, with no road crossings or other

facilities in between

- Packaging for Shipment E - Management and staff experienced in shipping SNM in all forms

Table 1 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE IDAHO NATIONAL ENGINEERING LABORATORY

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT INEL

Supporting Site Assets

- DOE Site Office E - Experienced in directing nuclear programs and large projects.

52 reactor projects have been completed at INEL.

- Sponsored and participated in meeting with GE Pu disposition team

- Site Contractor Management E - Operational experience managing nuclear processing and large

projects including many special purpose reactors

- Expressed strong support for use of electric power producing

fission reactor for Pu disposition and confidence that complex

could be built and operated successfully at INEL

- Considers plutonium to be a national resource

- Work Force - Technical and production oriented work force can support large

plutonium processing and reactor construction and operation project

- INEL has -12,500 employees of which "4600 hold professional degrees

- Utilities E - Adequate water supply and distribution system for reactor makeup

and complex operations

- Adequate electric power provided by Idaho Power 230 KV transmission

line to INEL to support large construction project

- Technology Development E - Argonne National Laboratory, West is technology center for DOE programs

_J - Quality Assurance E - Experienced in working to DOE quality assurance requirements! °

Also NRC and industry related (such as ASME, ANS, IEEE, NQA1 & 2, etc.)

"_] quality assurance requirements.

- Safety E - Management and staff experienced in meeting or exceeding all health

and safety requirements associated with nuclear programs. Long

history of safe operation at INEL.

- Environmental Protection E - Management and staff experienced in protection of environment.

Environmental restoration programs are operational at INEL. [TBR]

- Community Support E - INEL management believes there is broad community support in Idaho for

nuclear activities based on long history of safe operations at INEL

which will facilitate public acceptance of Pu Disposition Complex

- Idaho Falls community leaders under "Initiative 2000" will support

new nuclear activities at INEL which meet environmental concerns [TBR]

- President of Idaho State AFL-CIO supports present and future

projects at INEL

- Long history of support by Idaho elected congressional, state,

county, and city officials. Congressional delegations unanimously

supported defense-related projects such as Special Isotope Separation

Project (producing plutonium) and Complex 21 at INEL. New nuclearactivities WHICH DO NOT INCLUDE PERMANENT STORAGE OF NUCLEAR WASTE

SUCH AS SPENT FUEL will be welcomed. [TBR]

- Public interest groups include Environmental Defense Institute.

Is requesting large amount of information on radioactive and chemicalreleases and accidents and worker radiation records under FOIA.

DRAFT 15Jan93 file NDEPLOY.

************************ DRAFT TO BE REVIEWED WITH NTS ***********************

Table 2 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE

G = Greenfield. New Capabilities Required.

U = Upgrade of Existing Capabilities Possible.

E = Existing or Planned Capabilities Meet Requirements.

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS

Overall Site Qualification E - Currently operating. Has served for -40 Fears as DOE site for

nuclear weapons testing. Current mission is to maintain readiness

to resume nuclear testing if required.

- Occupies -1350 square miles -65 miles northwest of Las Vegas, NV.

Bordered on three sides by additional 4120 square miles federally

controlled (Nellis Air Force Range).

PuO2 Feed Material InterfaceI

(IF w/Nuclear Weapons Complex Reconfig Program}OO

- Pu Receiving and Storage G - New capability needed for handling Pu in forms for processing.

(within weapons program)

- Pu-to-PuO2 Conversion G - New capability needed for Pu-to-PuO2 conversion processing

(within weapons program}

- PuO2 Feed Interface Management U - Experience managing nuclear weapons testing could be extended

to capablility for management of interface with PuO2 supplier

Mox Fuel Fabrication

- MOX Pellet and Pin Fabrication G - New capability needed for MOX fuel fabrication processing

- ABWR MOX Assembly Fabr'ication G - NOX assemblies could be fabricated in greenfield facility

built for MOX pellet and pin fabrication

- Fresh MOX Assembly Storage G - Fresh MOX assemblies could be stored in greenfield facility

built for NOX pellet and pin fabrication

Tritium Target Fabrication Interface

(IF w/Nuclear Weapons Complex Reconfig Prog)

- ABWR Tritium Target Fabrication G - Target rods could be fabricated using components from commercial

(within weapons program) vendors in greenfield facility built for MOX pellet and pin fabrication

- Tritium Target Interface Management E - Experience managing procurement from commercial vendors is applicable

to management of interface with supplier of tritium targets whether

or not ABWR tritium targets are fabricated on-site

Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS

Reactor Plant Site Qualification

- Reactor Site E - 360 acre site identified for Tritium Supply Site within I0,000 acresfor Complex 21 available in east-central NTS. More if needed.

- Site Qualification U - Qualification of NTS for nuclear weapons testing activities isapplicable towards qualification as operating reactor site.

Has not participated in NPR or other reactor siting evaluations.

- Reactor Plant G - The ABWR plant will be a new facility

- Reactor Plant Operation G - Reactors have not been built at NTS. Reactor operation will hea new function.

Power to the Grid Interface

(IF w/IPP/Utility}

- Proximity of Commercial Grid E - Southern Nevada is hub for power transmission corridors connecting loadcenters and generation systems in Utah, California, Nevada, Arizona

s - Reactor-to-Commercial Grid Transmission G - New -100 mi transmission line and substation needed to link 1350 MWoreactor electrical output to 500 kV power transmission grid [TBR]

- Electric Power Sales G - Agreements needed- Potential value of baseload power is [TBD]

- IPP/Utility Interface Management U - Experience managing power distribution sytem for RTS applicableto management of interface with IPP/Utility. Extension of

capability needed to handle transmission and sale_ of 1350 MWe.

Spent Fuel Storage Interface

(IF w/Nuclear Waste Disposal Program)

- Spent Fuel Storage G - New capability needed for storage of spent fuel at NTS

(within waste disposal program)

- Spent Fuel Transport G - New capability needed for transport of spent fuel

{within waste disposal program)

- Spent Fuel Interface Management U - Extension of capability needed for management of interface withNuclear Waste Disposal Program whether or not spent fuel is

stored or transported by NTS

Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS

Waste Management

- High Level Waste (HLW) E - Management an4 staff experienced in handling HLW [TBR]

- Transuranic Waste (TRU} E - Management and 3tall experienced in handling TRU [TBR]

- Low Level Waste (LLW) E - Radioactive Waste Management Site in Area 5 includes 92 acres

for surface storage and disposal of LLW

- Hazardous Waste E - Management and staff experienced in handling hazarsous wastes [TBR]

- Mixed Waste E - Management and staff experienced in handling mixed wastes [TBR]

- Nuclear Waste Management Interface E - Experience managing nuclear wastes is applicable to management

of interface with Nuclear Waste Disposal Program

Tritium Recovery Interface

(IF w/Nuclear Weapons Complex Reconfig Prog)

- Extraction G - New capability required

(within weapons program)

- Purification G - New capability required

(within weapons program)

!- Reservoir Loading G - New capability required

(within weapons program)

- Tritium Recovery Interface Management U - Similar to spent fuel. Extension of capability needed for manageDent

of tritium recovery interface, whether or not some or all of the

tritium recovery functions are located at NTS

Safeguards & Security

- Accountability U - Management and staff experienced in strict accountability of

nuclear weapons compon-nts. Upgrading needed for accountability

of Pu in forms for processing.

- Protection U - NTS experience in protection of nuclear facilities is applicable to

providing additional personnel, fences, and guard posts for

Pu Disposition Complex

Safety and Environmental Approval

- Safety Approval U - Experienced with safety approval by DOE for weapons testing related

facilities. This capability will need to be upgraded for DOE and

NRC requirements for reactor and Pu processing facilities.

- Environmental Approval U - Management and staff have experience with Environlental Assessment andEnvironmental Impact Statement requirements. Currently preparing data

for EIS for upgrade alternative for Assembly/Disassembly function under

Nuclear Weapons Complex Reconfiguration Program. This exper_ Jnce is

applicable to extension of capability needed for reactor and

Pu processing facilities.

- Federal/State/Local Permitting E - Experienced in permitting at all levels

Table 2 {ton't} PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS

Transportation- Site Access E - Access by air via Las Vegas, NV. Desert Rock Airport on NTS has

7500' long x 100' wide runway capable of accepting jet aircraft.

- Access by road is four lane, divided US95 which intersects I15.Nature on-site road infrastructure. Adequate for construction.

- Access by rail via rail head at Las Vegas with Union Pacificline to NTS.

- Unloading & Inspection U - Management and stafC experience in receiving nuclear weaponscomponents is applicable to upgrading of capability needed for

receiving Pu in forms for processing

- On-Site Transport E -Nanagement and staff experienced in on-site transport of SNM- No public roads on NTS

- Packaging for Shipment E - Management and staff experienced in packaging nuclear componentsand wastes for shipment off-site

Supporting Site Assets- DOE Site office U - Experienced in directing nuclear weapons testing activities.

Applicable to extension of capability needed for reactor and

Pu processing activities.- Plutonium considered to be a resource

J

ho

- site Contractor Management U - Operational experience managing nuclear weapons testing activities.Applicable to extension of capability needed for reactor

and Pu processing activities.

- Work Force U - Trained, experienced, educated industrial base could supply skilled

professional, technical, craftspersons. Expansion needed for

Pu processing and reactor operation.

- NTS has -3500 employees

- Utilities E - "9E6 gpd water available at NTS from 14 existing wells.

-2E6 gpd beyond current usage available from Area 6 System for

proposed Complex 21 site. Additional wells can be drilled.

- Two independent 138 kV transmissions lines providing 25-30 NW peak

load, with 10-15 MW to be added within next few years.

Table 2 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE NEVADA TEST SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT NTS

Supporting Site Assets (Con't)

- Technology Development E - Nevada Research & Development Area (NRDA} experienced with technology

development for nuclear rocket program. ENAD facility located inthis area.

- Quality Assurance E - Experienced in working to DOE quality assurance requirements

- Safety and Health E - Management and staff experienced in meeting or exceeding all health

and safety requirements associated with nuclear programs

- Environmental Protection E - Management and staff experienced in protection of the environRent.

- Community Support E - Long-standing support from local cummunities for DOE defense prograls

- Congressional delegation, governor, and state, regional, and local

officials publicly support continued defense-related projects at NTS

- Nevada Test Site Contractors Association supports now programs at NTS

- Political opposition to plutonium storage at NTS. Connected with

opposition to disposal of high level waste at Yucca Mountain.

!

file ODEPLOY.

DRAFT 15Jan94 **************************** DRAFT TO BE REVIEWED WITH ORR ***********************

Table 3 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION

G = Greenfield. New Capabilities Required.

U = Upgrade of Existing Capabilities Possible.

E = Existing or Planned Capabilities Meet Requirements.

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR

Overall Site Qualification E - Has served for -50 years as DOE site for energy R&D and weapons uranaumand litiium production. Current missions are energy R_D, weapons

disaantling, and uranium and lithium storage, waste management,

and environmental restoration.

- Currently operating, with -[TBD] employees

- occupies -54 square miles "17 miles northwest of Knoxville, TN

- Three technology areas: Xl0 (Oak Ridge National Laboratory}, Yl2, K25

Pu Feed Material Interface

(IF w/Nuclear Weapons Complex Reconfig Program}

- Pu Receiving and Storage E - Management and staff experienced with SNM handling including Pu

(within weapons program} - Pu could be stored in [TBD] existing facilities

' U/G - Y12 facilities available for upgrading for Pu-to-PuO2 conversion.- Pu-to-PuO2 Conversion

t_ (within weapons program} New process equipment needed.- Also greenfield available within 10,000 acres proposed for Complex 21 site

just West of YI2

- Management and staff have lab scale Pu-to-PuO2 conversion experience [TBR]

- PuO2 Interface Management E - Extensive experience at ORR in management/handling/transport/accountabilitof enriched uranium, and also Pu and PuO2 [TBR], is applicable to

management of interface with supplier of PuO2 feed material whether ornot Pu-to-PuO2 conversion function is located at ORR

MOX Fuel Fabrication

- MOX Pellet and Pin Fabrication U/G - Y12 facilities available for upgrading and new process equipmentfor MOX blending, pellet mint.ring, and pin loading operations

- Also greenfield available within proposed Complex 21 site

Management and staff have pilot scale experience producing MOX powder

using sol-gel and gel sphere processes

- ABWR MOX Assembly Fabrication U - Existing Y12 facilities adaptable for MOX assembly fabrication [TBR]Could also be included within MOX pellet and pin fabrication facility

located at either Y12 or Complex 21 site

- Fresh MOX Assembly Storage U - Existing YI2 facilities adaptable for fresh MOX assembly storage [TBR]Could also be included within MOX pellet and pin fabrication facility

located at either YI2 or Complex 21 site

Table 3 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT eRR

Tritium Target Fabrication Interface

(IF w/Nucxlear Weapons Complex Reconfig Program)

- ABWR Tritium Target Fabrication U - Existing YI2 facilities adaptable with minimal refurbishing [TBR].

(within weapons program) New equipment for pin loading and closure needed.

- Could also be easily included within MOX pellet fabrication facility

located at either YI2 or Complex 21 site

- Management and staff experience fabricating fuel pins is applicable

to assembly of tritium targets using vendor supplied components [TBR]

- Tritium Target Interface Management E -- eRR has experience receiving and preparing tritium targets for

shipment to SRS under the production reactor program.

- Management and staff experienced with lithium handling, and with

coordination of procurement =tom commercial vendors, is applicable

to management of interface _°ch supplier of tritium targets whether

or not tritium targets are fabricated on-site

Reactor Plant Site Qualification

- Reactor Site E -- ABWR plant could be located within proposed Complex 21 site

- Former Clinch River Breeder Reactor (CRBR) site, owned by TVA,

is adjacent to eRR and could be utilized for ABWR

m - Site Qualification E - Data already compiled for Complex 21 site evaluation is applicable

to ABWR plant site qualification

- TVA site was characterized for CRBR plant

- Reactor Plant G - The ABWR plant will be a new facility

- Reactor Plant Operation E - err management and staff have operated several reactors. The new

reactor "Advanced Neutron Source" is planned by DOE at eRR.

Power to the Grid Interface

(IF w/IPP/Utility)

- Proximity of Commercial Grid E - eRR Central Control Facility at K25, formerly used as input

station for >2000 NWe from TVA, could be used as output stationfor 1350 HWe ABWR

- Reactor-to-Commercial Grid Transmission U - eRR has three primary 161 kV substations and eight primary

161 kV transmissions lines criss-crossing the site which could

be upgraded link with the 1350 HWe ABWR plant [TBR]

- Electric Power Sales U - Baseload power could be wheeled through the TVA grid for sale

to a nearby IPP/Utility. Agreements needed.

- The potential value of the baseload power is [TBD]

- IPP/Utility Interface Management E - Experience managing large amounts of electric power provided

to eRR in the past is applicable to management of the interface

with IPP/Utilities for power provided from eRR

Table 3 (Con't} PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR

Spent Fuel Storage Interface

(IF w/Nuclear Waste Disposal Program)

- Spent Fuel Storage E - Facilities at ORR for temporary storage, pending ultimate disposal,

(within waste disposal program) of spent fuel from the ABWR Pu Disposition Complex include _TBDJinclude [TBD]

- Spent Fuel Transport E - Management and staff experieced in transport of spent fuel

(within waste disposal program)

- Spen_ Fuel Interface Management E - Experience managing spent fuel from reactors at ORR is applicableto management of interface with US Spent Fuel Disposal Programwhether or not the spent fuel is stored at ORR

Waste Management

- High Level Waste (HLW) E - Management and staff experienced in handling HLW in the form ofirradiated reactor components such as control rods and core

structural elements |TBR]

- HLW is stored for future disposal at the [TBD] facilities at ORNL

- Transuranic Waste (TRU} E - Management and staff experienced in handling TRU- Solid transuranic waste is stored for future treatment and disposal

t_ at the [TBD] facility at ORNL. Solid Pu scrap from the NFS plant

t_ is also being transferred to ORNL. A new facility at ORNL to treat

and package solid transuranic waste to specifications for ultimate

disposal at WIPP has been proposed to DOE.

- Liquid and sludge transuranic waste is stored at the [TBD] storage

tank facility at ORNL. A new facility at ORNL to treat 3-600,000

gallons of transuranic liquid/sludge to specifications for ultimate

disposal at WIPP has been proposed to DOE.

- Low Level Waste (LLW) E - Management and staff experienced in handling LLW

- Hazardous Waste E - Management and staff experienced in handling hazardous wastes

- Mixed Waste E - Managemet and staff are experienced in handling mixed wastes- ORR has the only licensed operating Toxic Substances Control

Act (TSCA} incinerator in the US to handle mixed waste.

currently handling liguid. Capable of handling solid.

- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable tomanagement of interface with Nuclear Waste Disposal Program

Table 3 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION

CAPABILITY R_UIREMENT CURRENT CAPABILITY AT ORR

Tritium Recovery Interface

(IF w/Nuclear Weapons Compl _x Reconfig Program)

- Extraction G - A new facility is required

(within weapons program)

- Purification G - A new facility is required for purifying the tritium to specifications(withi_ weapons program) for shipment to the RTF at SRS

- Recovery G - A new facility is required at ORR. It is believed the Replacement

(within weapons program) Tritium Facility at SRS has been designated by DOE for tritium recovery

for the weapons program

- Tritium Recovery Interface Management E - The proposed Advanced Neutron Source reactor at ORR will

include a detriiation facility

- Experience managing in-pile experiments and tritium operations at ORR

is applicable to management of tritium recovery interface with the

Nuclear Weapons Complex Reconfiguration Program whether or not some or

all of the tritium recovery facilities are located at ORR

Safeguards _ Security

- Accountability E - Management and staff experienced in implementing strict

accountability procedures for storage, handling, and transport, of special nuclear materials in all forms including metal,

oxide components, and wasteO_

- Protection U - ORR has an existing security force, with experience in PIDAS zones,

which is applicable to providing additional personnel, fences, and

guard posts for protection of Pu Disposition Complex

Safety and Environmental Approval

- Safety Approval U - ORR has experience with the design requirements, safety analyses,

documents, safety reviews and audits, operationaA readiness reviews,

conduct of operations, Price-Anderson ammendmont rules, configuration

management, and quality assurance to obtain safety approval by DOEfor nuclear facilities including reactors. Since the DOE and NRC

requirements are similar, this capability at ORR can be upgraded toaccomodate NRC procedures.

- Environmental Approval U - ORR has experience with the environmental approval procedures which

result in the Record of Decision by DOE. These procedures are still

evolving, and it is anticipated that this capability at ORR will

need to be upgraded for new reactors and Pu processing facilities.

- Federal/State/Local Permitting E - ORR has experience with permitting at all levels

Table 3 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE OAK RIDGE RESERVATION

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT ORR

Transportation

- Site Access E - ORR has access to the site by road, rail, and barge which can

meet construction needs of the Pu Disposition complex

- Unloading • Inspection E - Management and staff experienced in receiving SNM in all forms

- On-Site Transport E - Management and staff experienced in on-site transport within

and between protected zones

- Packaging for Shipment E - Management and staff experienced in shipping SNN in all forms

Supporting Site Assets

- DOE Site Office E - Experience in directing nuclear prograEs and largo projects

is applicable to Pu Disposition Complex

- Site Conttractor Management E - Operational experience in management of plutonium and tritium

processing and technology and large projects including reactors

is applicable to Pu Disposition Complex

- Site contractor management has indicated strong interest in

the Pu Disposition Complex and believe ORR could accomodate

o the program

- Work Force E - Technical and production oriented work force can support large

plutonium and tritium construction and operation project

- utilities E - Adequate water for reactor makeup and complex operations

- Adequate electric power and other utilities to support

major construction project

- Technology Development E - ORNL is multipurpose research laboratory in energy related areas

- Quality Assurance E - Experienced in working to DOE quality assurance requirements

- Safety and Health E - Management and staff experienced in Beeting or exceeding all health

and safety requirements associated with nuclear programs

- Environmental Protection E - Management and staff experienced in protection of the environment.

Environmental restoration programs are operational at ORR.

- Conmunity Support E - Indications are that state and local officials will be receptive

to the handling of plutonium at ORR, based on recent public

hearings for the Complex 21 Program Environmental Impact Statement

file PDEPLOY.DRAFT 15Jan94

******************** DRAFT TO BE REVIEWED WITH PANTEX **********************

Table 4 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT

G = Greenfield. New Capabilities Required.

U = Upgrade of Existing Capabilities Possible.

E = Existing or Planned Capabilities Meet Requirements.

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX

Overall Site Qualification E - Currently operating. Has served for >40 years as DOE site for assemblyand disassembly of nuclear weapons. Current missions are if) fabricate

chemical high explosive components for nuclear weapons, (2) ass®ible

nuclear weapons for the nation's stockpile, {3) maintain and evaluate

nuclear weapons in the stockpile, and (4) disassolble nuclear weapons

being retired from the stockpile. Pantex has also been designated asan interim storage site for Pu pits.

- Occupies "16,000 acres -17 miles northeast of Amarillo, TX

PuO2 Feed Material Interface

(IF w/Nuclear Weapons Complex Reconfig Program)

!

tO - Pu Receiving and Storage G Handling of Pu in metallic form for processing would he new capability

Oo (within weapons program)

- Pu-to-PuO2 Conversion G - Management and staff experienced in handling weapons parts containing

(within weapons program) nuclear materials. Nuclear materials have not been processed in thepast, and this would be new capability for Pantex.

- PuO2 Feed Interface Management U - Experience managing nuclear weapons components could be extended tocapability for management of interface with PuO2 supplier

Mox Fuel Fabrication

- MOX Pellet and Pin Fabrication G - New capability needed for Pu processing for MOX fuel fabrication.- Pantex has experience with process robotics, including 70% automation

of glovebox operations and 100% automation of handling of the pit

in the Pit Reuse Program

- ABWR MOX Assembly Fabrication G - MOX assemblies could be fabricated in greenfield facilitybuilt for MOX pellet and pin fabrication

- Fresh MOX Assembly Storage G - Fresh MOX assemblies could be stored in greenfield facilitybuilt for MOX pellet and pin fabrication

Tritium Target Fabrication Interface

(IF w/Nuclear Weapons Complex Reconfig Prog)

- ABWR Tritium Target Fabrication G - Target rods could be fabricated using components from couercial

(within weapons program) vendors in greenfield facility built for MOX pellet and pin fabrication

- Tritium Target Fab Interface Management E - Experience managing procurement from commercial vendors is applicableto management of interface with supplier of tritium targets whether

or not ABWR tritium targets are fabricated on-site

Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX

Reactor Plant Site Qualification

- Reactor Site E - Land is available for a reactor. Specific potential sites have

not yet been selected.

- Site Qualification U - Characterization of Pantex site for nuclear weapons activities is

applicable towards qualification as an operating reactor site.

Has not participated in NPR or other reactor siting evaluations.

- Reactor Plant G - The ABWR plant will be a new facility

- Reactor Plant Operation G - Reactors have not been built at Pantex. Reactor operation willbe a new function.

Power to the Grid Interface

(IF w/IPP/Utility)

- Proximity to Commercial Grid [TBDJ

- Reactor-to-Commercial Grid Transmission |TBDJ

t_ - Electric Power Sales [TBD]!

- IPP/Utility Interface Management |TBD]

Spent Fuel Storage Interface

(IF w/Nuclear Waste Disposal Program)

- Spent Fuel Storage G - New capability needed for storage of spent fuel at Pantex

(within waste disposal program_

- Spent Fuel Transport G - New capability needed for transport of spent fuel

(within waste disposal program)

- Spent Fuel Interface Management U - Expansion of capability needed for management of interface with

Nuclear Waste Disposal Program whether or not spent fuel is

stored or transported by Pantex

Table 4 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX

Waste Management

- High Level Waste (HLW) G - HLW not currently generated or disposed of on-site

- Transuranic Waste (TRU) G - TRU not currently generated or disposed of on-site

- Low Level Waste (LLW) G - LLW not currently generated or disposed of on-site

- Hazardous Waste E - Management and staff experienced in handling hazardous wastes- Other than burning of high exploseve materials from disassembled

weapons, hazardous wastes not currently disposed of on-site

- Mixed Waste G - Mixed wastes not currently generated or disposed of on-site

- Nuclear Waste Interface Management U - Experience of Manager, Amarillo Area Office in TransuranicWaste Program is applicable towards expansion of capability

for management of interface of Pu processing and reactor waste

functions with Nuclear Waste Disposal Program

Tritium Recovezy InterfaceI

t_ (IF w/Nuclear Weapons Complex Reconfig Prog}

- Ext action G - New capability required

(within weapons program)

- Purification G - New capability required

(within weapons program)

- Reservoir Loading G - New capability required

(within weapons program}

- Tritium Recovery Interface Management U - Similar to spent fuel. Extension of capability needed for managementof tritium recovery i_terface, whether or not some or all of the

tritium recovery full. ions are located at Pantex

Safeguards & Security

- Accountability U - Management and staff experienced in strict accountability of nuclearweapons components. Safeguards & Security Directorate includes nuclearmaterial control, nuclear material accounting, and enhancement for

protection of SNN. Also security clearance and classified document

control. Expansion of capability needed for accountability of Pu in

forms for processing.

- Protection U - Pantex experience in protection of nuclear weapons facilities isapplicable to providing additional personnel, fences, and guard posts

for Pu Disposition Complex

- Safeguards & Security Directorate includes security force program.

Experienced with PIDAS zones.

Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT

CAPABILIT_ REQUIREMENT CURRENT CAPABILITY AT PANTEX

Safety and Environmental Approval

- Safety Approval U - Management and staff experience in safety approval by DOE for weaponsassenbly/disassembly facilities and operations is applicable to

extension of capability and procedures needed for DOE and NRC

requirements for Pu processing and reactor safety approval

- Environmental Approval U - Management and staff experienced in environmental permitting for weaponsassembly/disassembly facilities and operations is applicable to

extension of capability and procedures needed to include requirements

for Pu processing and reactor environmental approval

- Environmental Assessment for storage of pits without HE in progress

- FederalStateLocal Pernitting E - Experienced in permitting at all levels

Transportation

- Site Access E - Access by air via Amarillo, TX- Access by road via US-60, which intersects with 1-40 in Amarillo.

47 miles of paved roads on-site.

- Access by rail is [TBD]. 17 miles of railroad track on-site.

- Unloading & Inspection U - Management and staff experience in receiving nuclear weapons!

t_ components is applicable to extension of capability needed forreceivin_ Pu in forms for processing

- On-Site Transport E - Nanagement and staff experience in on-site transport of nuclearweapons components is applicable to transport of packaged Pu

- No public roads on Pantex site

- Packaging for Shipment U - Management and staff experience in packaging nuclear weaponscomponents for shipment off-site is applicable to extension of

capability needed for shipment of Pu and wastes in processing forms

- Experience includes use of Special Secure Transporter (SST) vehicles.

SST operation is directed from Pantex.

Supporting Site Assets- DOE Site Office E/U - Albuquerque Operations Office experience directing nuclear programs

and large projects is applicable to Pu Disposition Complex

- Amarillo Area Office experience in direction of nuclear weapons

assembly/disassembly related programs is applicable to extension

of capability needed for reactor operation and Pu processing

- Albuquerque Operations Office and Amarillo Area Office consider

plutonium to be a national resource

- Site Contractor Management U - Operational experience managing nuclear weapons assembly/disasselblyis applicable to extension of capability needed for reactor and

Pu processing

- Conmitted to growth by infusing new technologies and broadened base

of technical personnel into the Pantex Plant

Table 4 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE PANTEX PLANT

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT PANTEX

- Work Force U - "3000 employees. 29% with bachelor's degree or above. >50% ofPantex work force are technically skilled oxployees. But not

experienced with Pu processing and reactor operations.

- Utilities E - Adequate water available [TBR]- Adequate electric power for large construction project [TBR]

- Technology Development U - Advanced Technology Office (ATe) provides planning and implementationof new technologies from Battelle, DOE labs, industry, and universities

for activities related to nuclear weapons assembly and disassembly.

Expansion of this capability needed to support Pu processing and

reactor operation.

- Tester Design Engineering Department provides design, development,

modification, and maintenance of automated electronic measurement

systems for testing of nuclear weapons electrical ciruitry

- High Explosives Synthesis Facility develops new chemical processes

for explosives

- Other capabilities at Pantex applicable to technology development

include the Analytical Laboratory, Nondestructive Evaluation (XRay)

Department, Gas Analysis Lab, Explosives Test Site, and others

- Quality Assurance E - Experienced in working to DOE quality assurance requirements

- Safety and Health U - Management and staff experienced in meeting or exceeding all health, and safety requirements associated with nuclear programs

t_ - Experience in safety for weapons assembly/disassembly facilities

and operations is applicable to extension of capability needed

for Pu processing and reactor operation such as criticality

alarm and Pu detection and containment

- Environmental Protection U - Management and staff experienced in protection of environment.Extension of capability needed for Pu prccessing and reactor operation.

- Community Support U - Community Relations Department activities include Information andAwareness, Community Involvement, Employee Communications, and

Educational Development

- Local community leaders in Panhandle 2000 support continuing

current nuclear weapons activities at Pantex

- Pantex is largest employer in Amarillo area

- State political leaders concerned about long term storage of pits

and potential for contamination of Ogalalla aquifer by Pu processingactivities

- Ranching and farming interests concerned about any adverse impact

on source of water

- University of Texas, Austin, Department of Economic Geology conducting

study of geologic structure to conduct water and potential for

contamination. Considers plutioium a resource.

- Public interest groups active around Pantex include Panhandle Area

Neighbors and Landowners (PANAL), Serious Texans Against Nuclear

Dumping (STAND}, Save Texas Agricultural Resources (STAR), andPeace Farm

DRAFT 15Jan94 file SDEPLOY.

************************* DRAFT TO BE REVIEWED WITH SRS *************************

Table 5 PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

G = Greenfield. New Capabilities Required.

U = Upgrade of Existing Capabilities Possible.

E = Existing or Planned Capabilities Meet Requirements.

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Overall Site Qualification E - Has served for >40 years as a DOE site for production and processingof plutonium and tritium. Current missions include processing and

storage of fissile materials, tritium processing, waste management,and environmental restoration.

- Currently operating, with -[TBD] employees

- Occupies -300 square miles adjoining the Savannah River -13 miles

southeast of Augusta, GA and -12 miles south of Aiken, SC

PuO2 Feed Material Interface

(IF w/Nuclear Weapons Complex Reconfig Program)

- Pu Recieving and Storage E - Plutonium Storage Facility (PSF} in SRS Separations Area was

(within weapons program) built to receive, store, monitor, retrieve, and sh_p packaged Pu,

and is adaptable to reveiving and storage of Pu metal from a weapons

site for conversion to PuO2 for Pu disposition

- Pu-to-PuO2 Conversion E - New Special Recovery Facility (NSR) in SRS Separations Area was

o (within weapons program) built to convert impure Pu metal or oxide from scrap to pure

t_ Pu nitrate solution, and is adaptable to producing PuO2 to spec

with some refurbishing

- PuO2 Feed Interface Management E - Experience in management/handlingtransport/accountability of

plutonium is applicable to management of interface with supplierof PuC2 feed material whether or not Pu-to-PuO2 conversion is

located at SRS

MOX Fuel Fabrication

- MOX Pellet and Pin Fabrication U/G -- Waste Tank Equipment Gallery (WETG} at Barnwell site can be

upgraded to house multi-story, automated fabrication facility.

New equipment for powder blending and sintering and pin

i_ading required.

- Greenfield sites also available well within SRS boundaries

- Management and staff experienced in Pu processing

- SRS has process automation development capability applicable

to design of MOX fabrication facility

- ABWR MOX Assembly Fabrication U - Could be co-located with either pellet/pin fabrication or fresh

assembly storage. New equipment for pin loading and closure needed.

- Fresh _OX Assembly Storage E - Main Processing Facility at Barnwell site has space for

temporary storage of assemblies until transport to reactor

- Storage space available at existing, not-currently-operating

reactors could meet requirements with some refurbishing

Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Tritium Target Fabrication Interface

(IF w/Nuclear Weapons Complex Reconfig Program)

- ABWR Tritium Target Fabrication E - Production Reactor Fuel and Tritium Target Fabrication Facility in300M Area adaptable for assembly from vendor supplied components

(within weapons program) with some refurbishing

- Could also be located within MOX fabrication facility

- Tritium Target Fab Interface Management E -- Experience managing fabrication of tritium targets for productionreactors and in procurement from commercial vendors is applicable

to management of interface with supplier of tritium targets whether

or not ABWR tritium targets are fabricated on-site

Reactor Plant Site Qualification

- Reactor Site E - ABWR plant could be located on site selected for NPR- Alternate NPR sites at SRS also available

- Site also available at Vogle Station, owned by Georgia

Power, just across Savannah River from SRS

- Site Qualification E - Site data already compiled for NPR and Complex 21 evaluationsis applicable to ABWR plant site qualification

- Site at Vogle Station is qualified for reactor

- Reactor Plant G - The ABWR plant will be a new facility!

t_

- Reactor Plant Operation E - Management and staff experienced in reactor operation.K Reactor, last of production reactors, operational until 1992.

Currently in cold standby readiness to restart if directed.

Power to the Grid Interface

(IF w/IPP/Utility)

- Proxlmity of Commercial Grid E - Commercial grid available through current connection to SCE&G lines- Commercial grid also available at Vogle Station, owned by Georgia Power,

just across Savannah River from SRS

- Reactor-to-Commercial Grid Transmission U - Existing high voltage transmission lines crossing SRS need to beupgraded to link with ABWR plant site

- Electric Power Sales U - Power could be brokered through existing DOE organization,Southeastern Electric Power Authority, but changes in legislation

and customer base required to sell power from nuclear plant

to customers with baseload needs.

- Power could he wheeled through SCEaG grid or from Vogle Station

through commercial grid for sale to a nearby IPP/Utility.

- At present, SRS power interface is with SCE&G, not Georgia Power

- The potential value of the baseload power is [TBD]

- Power sales agreements needed

- IPP/Utility Interface Management E - Experience in managing large amounts of electric power providedto SRS in the past is applicable to management of interface

with the IPP/Utilities for power provided from SRS

Table 5 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Spent Fuel Storage Interface

(IF w/Nuclear Waste Disposal Program)

- Spent Fuel Storage E - P-, K-, and L-Reactor pools now being used to store U/A1 alloy fuel

(within waste disposal program} from production reactors. A proposed plan is to process this fuel

in F-Canyon. Should take -3 years once started. Then pools could

be available for spent ABWR MOX fuel.

- C-Reactor pool currently filled with water and empty of fuel.

Can be used for spent fuel with minimal upgrading to meet

water purification system requirements

- R-Reactor pool is dry and empty

- Main Processing Facility at Barnwell site has large pools

currently dry and eRpty and available. Barnwell County Council

has Economic Development Initiative for "Center for Nuclear

Materials Management", which includes temporary storage of

spent commercial LWR fuel.

- Spent Fuel Transport E - Management and staff experienced in spent fuel transport

(within waste disposal program)

- Spent Fuel Interface Management E - Experience managing spent fuel from production reactors is

applicable to management of interface with Nuclear Waste Disposal• Program whether or not spent fuel stored at SRS

!

t_ Waste Managementt_

- High Level Waste (HLW) E - Management and staff experienced in handling high level waste

(including transuranics} in the forms of solid and liquid from

irradiated fuel reprocessing and irradiated reactor components suchas control rods and core structural eleRents

- Liquid/sludge high level wastes (including transuranics} from

reprocessing operations stored at tank farms (241-F and 241-H)

- Construction of Defense Waste Processing Facility (DWPF} is

complete and currently in cold test mode. Will vitrify high

level liquid/sludge waste (including transuranics} into

glass logs, to be stored on-site until transport to Yucca

Mountain for disposal.

- Transuranic Waste (TRU} E - Management and staff experienced in handling transuranic wastes

- Solid transuranic wastes stored at Solid Waste Disposal Facility

- Liquid/sludge transuranic wastes considered to be high levelwaste and treated as noted above

- Low Level Waste (LLW) E - Management and staff experienced in handling low level wastes

- SRS has Solid Waste Disposal Facility for solid waste including

used equipment

Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Waste Management (Con't)

- Hazardous Waste E - Management and staff experienced in handling hazardous wastes

- Mixed Waste E - Management and staff experienced in handling mixed wastes

- SRS has broken ground fen Consolidated Incineration Facility to

dispose of burnable wastes from all site operations

- Nuclear Waste Interface Management E - Experience managing all forms of nuclear waste is applicable to

management of interface with Nuclear Waste Disposal Program

Tritium Recovery Interface

(IF w/Nuclear Weapons complex Reconfig Program)

- Extraction U - Existing facility for tritium extraction from _ tritium

(within weapons program} targets needs upgrading for extraction from ABWR targets

- Purification E/U - Existing tritium purification capability could moot

(within weapons program) requirements, but will soon need replacement duo to age.

Upgrade of Tritium Extraction/Separation Facility (232-H}

has been proposed to DOE.

- Reservoir Loading E - Replacement Tritium Facility (RTF) is now facility!

t_ (within weapons program)

- Tritium Recovery Interface Management E - Experience managing tritium operations is applicable to

managing the tritium recovery interface whether or not some

or all of the tritium recovery facilities are located at SRS

Table 5 (ton't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Safeguards & Security

- Accountability E - Management and staff experienced in implementing strict accountability

procedures for storage, handling, processing, and transport of special

nuclear materials in all forms including metal, oxide, componentsand waste

- Protection U - SRS has existing security personnel, with experience in PIDAS zones,

which can provide protection for Pu disposition facilities with

some upgrades for additional personnel, fences, and guard posts

depending on actual location on-site

Safety and Environmental Approval

- Safety Approval U - Management and staff experienced with requirements, analyses,

documents, and reviews required to obtain safety approval by DOE

for nuclear facilities including reactors. Since DOE and NRC

requirements are similar, this capability can be upgraded to

accomodate NRC requirements.

- Safety approval experience for K-Reactor and NPR programs

applicable to Pu Disposition Complex

- Environmental Approval E - Management and staff experienced with environmental approval

requiremments to obtain Record of Decision by DOE with EPA, approval for nuclear facilities including reactors.

t_ - SRS has established approach of detailed procedures for evaluations,

_J documents, and reviews for each stage of approval processes

- Site data developed for NPR and Complex 21 evaluations applicable

to Pu Disposition Complex.

- Federal/State/Local Permitting E - SRS has pro-active approach to permitting at all levels

Transportation

- Site Access E - SRS has access to site by road, rail, and barge docking facilities,

which can meet construction needs of Pu Disposition Project with

minor upgrading. Savannah River can be made navigable for barge

transport of heavy equipment.

- Unloading & Inspection E - Management and staff experienced in receiving SHlq in all forms

- On-Site Transport E - Management and staff experienced in on-site transport of SNM

in all forms within and between protected zones- Public road crosses SRS

- Packaging for Shipment E - Management and staff experienced in shipping SNlq in all forms

Table 5 (Con't) PLUTONIUM DISPOSITION COMPLEX DEPLOYMENT CAPABILITIES AT THE SAVANNAH RIVER SITE

CAPABILITY REQUIREMENT CURRENT CAPABILITY AT SRS

Supporting Site Assets- DOE Site office E - Experienced in directing nuclear programs and large projects

- Have sponsored tours and presentations to expedite independent

review and evaluation of SRS capabilities for Pu disposition

by GE team

- Have expressed interest in status of Pu Disposition Study

independent of site and reactor type

- Site Contractor Management E - Operational experience managing plutonium and tritium processingand technology and largo projects including production reactors

- Site contractor management has indicated strong interest in

the Pu Disposition Complex and believes SRS could accolodate

the program

- Work Force E - Technical and production oriented work force can support largeplutonium and tritium processing and reactor construction and

operation project

- Utilities E - Adequate water supply and distribution system for reactor makeupand complex operations. Three pumping stations on Savannah River

can deliver 3E6 gpm to SRS.!

t_ - Adequate electric power to support large construction projectOo

- Technology Development E - Savannah River Technology Center (SRTC) at SRS devoted to solvingproduction problems and improving processes of nuclear materials

handling, storage, processing, reactor operation, and waste management

- Quality Assurance E - Experienced in working to DOE quality assurance requirements

- Safety and Health E - Management and staff experienced in meeting or exceeding all healthand safety requirements associated with nuclear programs

- Environmental Protection E - Management and staff experienced in protection of the environment.Environmental restoration programs operational at SRS.

- Community Support E - SRS management believes indications are that state and local officialswill be receptive to new Pu processing and reactor operations of

Pu Disposition Complex, based on response to NPR and Complex 21

proposals- SRS management has found widespread local community support for

operations at the site. Active program of communications with

community leaders and public is maintained.

- Barnvell County Economic Development Initiative includes proposal

for "Center for Nuclear Materials Management" which is to be a

private, commercial engineering laboratory working on ways to

manage waste

- Energy Research Foundation (ERF), located in Col_L_bia SC, is loading

public interest group. Concerns include long torn storage of

nuclear materials and release of tritium contaminated waste.

4.4.1 TRANSPORTATION LOGISTICS FORTRITIUM PRODUCTION

4.4.1.1 Summary

Tritium production tnvolves transportation of enrtched ltthtum andtritium, both

of which are classified as "Other Nuclear HaterJal" and require appropriate

safeguards, security and accountability measures. In addition, the irradiated

target rods must be shippedto the extraction facility and the spent target rods

(after tritium extraction) shipped to the waste dtsposal site. The results of

this evaluation confirmed that the existing and/or planned transportation

infrastructureshouldbe adequatetomeetthetritiumproductionneeds.Further,

relativelyfew shipmentswill be required.Thus,transportationlogisticsare

not likelyto be a controllingconsiderationin sitingthe tritiumproduction

facilities.

4.4.1.2 Background

The purposeof thistaskwas to assessthe transportationlogisticsassociated

with tritiumproductionin the ABWR in the contextof applicablerequirements,

the existingand planned infrastructureand the deploymentoptions under

consideration.The scopeof the assessmentincludedtransportationrelatedto

targetfabrication,shipmentof freshand irradiatedtargetrods,the tritium

productand the spenttargetrodsafterextractionof the tritium.

As discussedin previoussectionsof thisreportthe referenceplanfortritium

productionin the ABWRconsistsof:

• Commercialfabricationof alltargetrodcomponentsincludingtheenriched

LiAl02pellets. The natureand locationof thesefabricationfacilities

was purposelyleftopento maximizeDOE flexibility.,

• Assemblyof the targetrods in a fuel fabricationfacility,but not

necessarilythe MOX fuelfabricationplant.

4.4.1-I

* Loading the target rods in the fuel bundle at the MOXfuel plant or thereactor.

• Discharge of all target rods after a one cycle exposure.

• Removalof the irradiated target rods from the fuel bundle in the reactor

spent fuel pool for shipment to the extraction facility.

• Extractionof tritiumfromthe irradiatedtargetrods in a new facility

located adjacentto the existingSavannahRiver Site (SRS) tritium

facilities.

• Disposal of the spent target rods after extraction as LowLevel Waste at

a DOEor commercial disposal site.

Alternatives considered which could impact the transportation logistics for

tritium production include:

• Fabrication of the LiAI02 pellets and/or assemblyof the target rods in a

facilityor facilitiesco-locatedwiththe reactor.

• Irradiationof the targetrodsfor multiplecycles.

• Co-locationof the extractionfacilitywith the reactor.

• Disposalof the spenttargetrodsafterextractionas corecomponentsin

the federalrepositoryfor spentnuclearfuel.

4.4.1.3 Discussion

The key factorswhich influencethe transportationlogisticsfor tritium

productionare the numberof targetrods irradiatedper year, theirdesign,

enrichedlithiumand tritiumcontentandthe activitylevelof the spenttarget

4.4.1-2

rods. Another potentially important factor is the extent to which the tritium

facilities are co-located with each other or with other related facilities.

The reference ABWRcore design for tritium production contains 872 fuel

assemblies with four enriched ltthtum target rods per assembly. The target rods

have the same external dimensions as the ABWRfuel rods. Each fresh target rod

contains approximately 20 grams of Li e. The average trttium content of a target

rod at discharge after one cycle exposure is approximately 104 curies or about

I gram. The gammadose rate from a spent target rod, one year after discharge

is expected to be on the order of 100 R/hr at a foot.

(_ontro_and Accountabilitv

Both enriched lithium and tritium are classified as "Other Nuclear Material"

under DOE Order 5633.3A, Control and Accountability of Nuclear Material.

Enrichedlithiumis classifiedas CategoryIV,AttractivenessLevel E,which must

have material control and accountabilityfor quantitiesof I kg or mo-e of Lia

(-50 target rods). Since the reference plan involves irradiationof ~3,500

target rods per year the transportationof the enriched lithium for pellet

fabrication,the finishedLiAL02pelletsand the fresh target rods will require

appropriatecontrol,accountabilityand safeguards. In the event that shipment

of the LiAl02pelletsor finishedtargetrods is consideredto involveclassified

configurationor contentthe physical protectionprovisionsof DOE Order 5632.5

would also apply.

The spent target rods, which containabout 85% of their originalLie inventory,

are plannedto be transportedin bundlesof 49 rods. In this configurationthey

will exceed the 100 rem/hr at 3 feet criteriaof 10 CFR 67 and 73.60 and can be

consideredas self protecting.

Under DOE Order 5633.3Atritiummust havematerialcontroland accountabilityfor

quantiLiesof 0.01 gram or more (~100 curies). Quantitiesof tritium greater

than t. grams are classifieda Category Ill,which requiresadditionalcontrol,

accountabilityand safeguards. While individualshipping containersmay have

4.4.1-3

less than 50 grams, the total quantity in a shipment wtll undoubtedly exceed this

limit in order to keep the number of shipments to a reasonable level.

A spent target rod after extraction is expected to contain about 50 curies of

residual tritium. The target rod handling concept for shipping, extraction and

waste disposal is to keep them in a 7x7 bundle which has the same external

dimensions as a BWRfuel assembly. This maintains compatibility with existing

and planned transportation systems. Thus, a spent target waste package would

contain -2,500 curies of residual tritium. While itwt11 be necessary to account

for this residual tritium from a materiai balance standpoint, the waste package

_i11 still have a dose rate that exceeds the self protection criteria of 10 CFR

67 and 73.60.

Safeguards and security provisions for Category III and IV materials require

development of protection provisions to be contained within a site-specific

safeguardsand security plan and/or a Master Safeguardsand SecurityAgreement

(MSSAs). Provisionsrequiredto be in thisdocumentsare identifiedin DOE Order

5632.2A.

A Material Controland Accountability(MC&A)Plan will be also requiredfor the

enriched lithiumand tritium. The level of controland accountabilitywould be

consistent with the economic and strategic value of these materials. An

implementationguide for DOE Order 5633.3A has been prepared which describes

methods for meeting requirementsof this order. Since a MSSA and a MC&A plan

will also be required for the ABWR f,aeland it is expected that the control,

accountabilityand safeguardsrequirementsfor enrichedlithiumand tritiumwould

be incorporatedin these documents.

In additionto DOE Order 5632.2Aand 5633.3A,the provisionsof DOE Order 5633.4,

Nuclear Materials Transactions: Documentation and Reporting, and DOE Order

5633.5, Nuclear Materials Reporting and Data Submission Procedures,are also

applicable and provide additional detail relating to the requirements.

Requirementsfor the scope and contentof MC&A plans are to be determinedby the

Manager, DOE Field Office.

4.4.1-4

Shipmentof CategoryIII quantitiesof tritiumcan be made by SST, or by

government-ownedtruck, exclusive-usecommercialcarrieror rail with the

appropriatecontrols(see Section7 for details). Enrichedlithium,being

CategoryIV,canalsobe shippedbyDOEapprovedcommercialcarriersandinvolves

lesscontrolsthan CategoryIII quantitiesof tritium.

)hiDm_ntof EnrichedLithium.LiAlO2_Pelletsand FreshTaraetRods

DiscussionswithDOE staffat the SavannahRiverSite (SRS)indicatethatoff-

siteshipmentof enrichedlithiumhasbeenveryinfrequent.Shipmentsthathave

takenplacehave involvedshipmentof lithiumin oxideform. The materialhas

beenplacedin a containerand sealedin an inertatmosphere.The containerhas

then beenplacedinsidea metaldrum for shipmentby a commercialcarrierthat

is bondedand certifiedby DOE. Theywere unawareof any near-termchangesin

regulations.

The quantitiesof enrichedlithiumand LiAl02pelletsneededby the ABWR are

smallenoughthattheycouldeachbe accommodatedin a singleannualshipmentif

desired.Theunirradiatedtargetrodswouldbe shippedincontainerssimilarto

thosecurrentlyusedforfreshBWRfuel. Thiswouldrequire25 to 30 containers

to shipthe 3,500targetrodsneededeachyearwhichcouldbe accommodatedintwo

shipments.Becauseof the relativelyfewshipmentsinvolvedit isunlikelythat

transportationwill be an importantconsiderationin sitingthe LiAI02pellet

fabricationor targetrod assemblyoperations.

_hiDmentof Irradiated Target Rodsto the Extraction Facilit,y

As indicated in Section 4.4.7.2 the reference plan is to ship the irradiated

target rods to a newextraction facility at the SRSfor recovery of the tritium

product. An alternate considered was to perform the extraction and initial

purification steps in a facility co-located with the reactor andship the tritium

to SRSfor final processing and storage. In either case the target rods wouldbe removedfrom the fuel assemblies in the reactor spent fuel pool and loaded

4.4.1-5

into transport canisters. These transport containers would hold 49 target rodsand have the sameexternal dimensionsandconfiguration as a BWRfuel assembly.

There are a numberof commerciallyavailable, licensed spent fuel shipping casks

that can be used to transportthe irradiated target rods for e_ther on-site or

off-site shipment. Theserange from a 23 ton, two assembly, l(_gal-weight truckcask to a 70 ton, eighteen subassemblyrail cask. The other option is to design

a special purposecaskwhich could undoubtedlyaccommodatelarger target loadings

than a spent fuel cask becauseof the reduced decay heat and radiation levels.However, since commercial cask shipment is relatively inexpensive (four rail

shipments per year at $4k per day) the investment in a special purpose cask for

target rods maynot be warranted.

Shipmentof Tritium

DOE has two typesof containersfor shippingtritiumas describedbelow. Both

containersarecurrentlyconfiguredto shipthe tritiumas a gas. WhileDOE is

consideringchangingtheformof tritiumshipmentfromgas to a metalhydridethe

capacityof the shippingpackagesis expectedto remainessentiallythe same.

The UC-609shippingpackage(Figure4.4.7-I)provides'containmentand offers

impactand thermalresistancefor shipmentscontainingtritium,in any of its

forms,duringtransportunderbothnormaland accidentconditions.The tritium

to be shippedisplacedwithinanappropriatestoragevessel.The storagevessel

isplacedwithina stainlesssteelcontainmentvesselandthecontainmentvessel

placedwithinan insulatedsteeldrum.

The internalcavityavailableforinstallationof a tritiumstoragevesselis 10

inchesin diameterby 31 incheslong. The UC-60gshippingpackagecan contain

30 gm-molesgas,not morethan25 gm-molesof whichmay be tritium(150gramsor

1.5millioncuries).The containmentvesselissurroundedby aminimumof 2-3/4

inchesof Celotexinsulation.A steeldrum surroundsthe insulationand is the

externalsurfaceof the package. The packageis 25 inchesin diameterand 55

incheshighandweightsa maximumof 500pounds.

4.4.1-6

The containment vessel is considered to be the primary containmentboundaryand

wtll contain the tritium if the packageis exposedto the normal or hypothetical

accident conditions specified tn 10 CFR71. The storage vessel receives no

credit for trttium containment but is to be designed, certified and tested to

provide the maximumassuranceof containment under all shipping conditions.

Tritium is shipped in the LP-50 Packageat low pressure (23.2 psia at 25 C) in

a 50 liter 304L stainless steel container surroundedby an aluminumvessel and

Celotex insulation at least 4 inches thick in a ]6-gauge steel drumwhich is 23.5

inches ODby 40 inches. This packagehas a maximumcapacity of ]93,500 curies

of tritium but current shipments are being limited to about ]25,000 curies oftritium.

The LP-50is expectedto be removedfromservicein about3-4yearsandwill be

replacedby a containercurrentlyunderdevelopment.Thisreplacementshipping

containeris expectedto utilizeuranium-hydrideand have a tritiumcapacity

comparableto the LP-50.

Basedon the size,weightandcapacityof theUC-609shippingcontainer(orits

replacement)it ispossibleto fitthecontractquantityof tritiumintoa single

shipment. However,prudentpracticewouldargueagainstcommittingthe entire

annualoutputto a singleshipment.

Shipmentof SpentTarqetRodsAfterExtractionof the Tritium

Spenttargetrodsareplacedin a canisterwiththe sameexternalconfiguration

as a BWR fuelassemblyandwouldbe compatiblewithbothexistinglicensedspent

fuelcasksandthe transportationsystemplannedfordisposalof commercialLWR

spentfuel. As a result,the transportationlogisticsw_ll be determinedby

whetherthe spenttargetrods are disposedof as low levelwasteat a DOE or

commercialdisposalsiteor as corecomponentsinthe federalrepository.Under

currentregulationseitheroptionis possible.

4.4.1-7

Plywood disk_j_

Heat shield - Plasticplug

Cerablanket insulation

(_ !' - _

Aluminum honeycomb, .... 5.0 in. on each end

3.7 in. on sides

" " Carbon steel drum

\ \ 24 in. insidediameter52½ in. insideheight

_ , / / Rubber I_ads

Vesselcarrier

\\ X

\\

54.5 in. Storage vesseltotal height

\\

I

Containment vesselType-316 stainlesssteel

\ \ _ 18 in. outside diameter1/8-in.-thick wall

Cavity 10 in. diam x 31 in. long

-,j Aluminum tube

\\

Celotex insulation/ / 2.9 in. on sides

4.0 in. on each.snd

l Cavity 18in. diam x 44 in. longI= 25 in. max diam

Figure 4.4.1.1

Model UC-609 Shipping Package

4.4.1-8

If the spent target rods are disposed of as low level waste the transfer cask

must be compatiblewith the packagingand unloadingrequirementsat the disposal

site. Further,unlessthe extractionsite is also permittedfor low level waste

disposal,the cask would have to be licensed for use over public roads. Whi_.q

it may be possible to interfacea commercialspent fuel cask with the disposal

site packagingand unloadingrequirementsa specialpurposecask is likelyto be

a more practicaland cost effectivesolutionin this case.

If the spent target rods are disposed of a core components in the federal

repositorythe followingconsiderationswould apply. A morecompletedescription

of this transportationsystemwas provided in Section4.4.4.

I0 CFR 961 providescriteriafor "standardnuclearfuel" which can be shippedto

the federal repository. This regulationalso allows for shipment of non-fuel

componentssuch as controlspiders,burnablepoisonrod assemblies,controlrod

elements, thimble plugs, fission chambers, and primary and secondary neutron

sources,that are an integralpart of the fuel assemblyand which do not require

special handling. While not directly applicable to spent target rods it is

likelythat they could be classifiedas "core components"if DOE chose to do so.

A legal-weighttruck cask is available for shipment of spent nuclear fuel

assemblies. This cask can hold g canistersof target rods which would require

eight annualshipmentsfromthe extractionfacilityto the MonitoredRetrievable

Storage (MRS).

Two Multi-PurposeCanistershave been conceptuallydesigned under the Civilian

RadioactiveWaste Management System (CRWMS) for rail transportationof spent

nuclearfuel from civilian reactorsites to the MRS or the FederalRepository.

The smaller MPC will hold 24 BWR assembliesand therefor would require three

annual shipmentsof consolidatedtarget rods. The largerMPC would accommodate

40 canisters of consolidatedtarget rods. Two annual rail shipmentswould be

requiredwith this larger MPC for the referencetritiumproductioncycle.

4.4.1-9

Other Considerations

Although the reference design for tritium production involves discharging all the

target rods after one cycle the potential exists to increase target rod exposure

to three cycles. This would reduce shipptng requirements throughout the tritium

cycle. However, since the number of shipments with one cycle exposure is

relatively small a change to three cycle exposure would not substantially change

the transportationlogisticsfor tritiumproduction.

Co-location of various tritium production operations can also reduce the

transportationrequired. While there are other good reasons to consider co-

location,transportationis not likely to be a controllingfactorbecauseof the

relativelyfew shipmentsinvolvedeven with a totallydispersedconfiguration.

From a transportationviewpoint,the differencein location of the extraction

facility (SRSor the reactorsite) affectsthe numberof shipmentsof irradiated

target rods and how the tritiumproduct is transportedto its end use location.

Locating the extraction facility at SRS avoids shipping the tritium in an

attractiveform (e.g purifiedproductgas or metal hydride). However, it could

requirean additionallongdistanceshipmentof the irradiatedtargetrods unless

they were disposedof as Low Level Waste at the SRS site. Again, becauseof the

existinginfrastructureand the few shipmentsneeded,transportationis unlikely

to be a controllingconsiderationin siting the extractionfacility.

The transportationrequiredto disposeof the spent target rods after extraction

is dependenton the waste disposalmethod selected (Low Level Waste or Federal

Repository). Disposalas a low level waste would probablybe less expensiveand

does not competewith spent fuel for priority in the transportationsystem or

space in the federal repository.

Under existing regulationspertaining to the Federal Repository the highest

transportationand disposalpriorityis assignedto the oldestspent fuel. Since

the MRS/Repositorymay just have begun operationat the time spent target rods

would need to be shipped, expanded storage would probably be needed at the

extractionfacility if disposal in the FederalRepositorywere selected.

4.4.]-I0

4.4.1.4 References

DOE Order 5630.13A,Master Safeguardsand SecurityAgreements

DOE Order 5630.14A,Safeguardsand SecurityProgram Planning

DOE Order 5632.2A, Physical Protectionof Special Nuclear Material and Vital

Equipment

DOE Order 5633.2A, Control and Accountability of Nuclear Materials:

Responsibilitiesand Authorities

DOE Order 5633.3A,Control and Accountabilityof NuclearMaterials

DOE Order 5633.4,NuclearMaterialsTransactions:Documentationand Reporting

DOE Order 5633.5, NuclearMaterialsReportingand Data SubmissionProcedures

Certificateof ComplianceUSA/6678/B()(DOE)Rev. O, for the LP-50 ShippingCask

Certificateof ComplianceUSA/9932/B(U)(DOE) Rev. 4, for the UC-609Shipping

Package

DPSPU-74-124-5,SafetyAnalysisReport:PackageLP-50TritiumPackage,May 1975:

Rev 2 issuedApril, 1988

Officeof CivilianRadioactiveWasteManagement(OCRWM)Bulletin,MPC Conceptual

Design Report Submittedto OCRWM, November,1993.

UCRL-52424,SafetyAnalysisReporton Model UC-609ShippingPackage,August 1977

UCRL-ID-111494,SafetyAnalysisReporton Model UC-609B(U) DOE ShippingPackage

4.4.1-II

4.4.2 TRANSPORTATION OF PLUTONIUM MATERIALS TO MOX

FABRICATION FACILITY

A. Off-Site Shipments

One of the feed materials for the mixed oxide fabrication plant is plutonium oxide. The source of

the plutonium oxide is the plutonium metal pits from disassembled nuclear weapons. Present

reconfiguration plans for the DOE weapons complex envision the first module of a plutonium

storage facility to be built by 2001 and a complete plutonium storage and processing facility to be

in operation by 2010. The nature of the stored plutonium (whether it will be metal or oxide) has

not been determined. There are severe problems in storing plutonium oxide for long periods of

time including americium buildup, gas evolution, and the hygroscopic nature of the powder.

Recommendations have been made that Pu02 should not be stored for more than three years.

Whatever storage for is selected, it appears that the source of material for the plutonium

disposition program will be from these new storage facilities.

While the conversion of the plutonium metal to oxide is not specifically addressed in the DOE

Phase 1C Statement of work, it is addressed here to provide a more complete description of

program needs. This conversion could be performed by DOE contractors at the fuel fabrication

site or a separate site. A discussion of the transport of plutonium metal forms follows, as well as

a description of the transport of plutonium oxide to the fuel fabrication plant.

Plul_onium M¢l_tl Transport

The shipment of plutonium metal pits to various DOE facilities has been safely carried out for

years. Since these are U.S. Government shipments and consist of nuclear weapons components,

they are regulated under DOE Order 5610.1, "Packaging and Transporting of Nuclear Explosives,

Nuclear Components, and Special Assemblies." The order states that these materials must be

packaged and transported to provide a level of safety at least comparable to that provided by

packaging and shipment, in accordance with applicable regulations of other radioactive and

explosive material. The references cited are DOT 49 CFR 100 through 179, "Hazardous Materials

Regulations." (These in turn, refer to Title 10 CFR 71, "Packaging of Radioactive Material for

Transport" which promulgates Federal regulations for the packaging of radioactive material for

4.4.2-1

transport); DOT 49 CFR 390-397, "Federal Motor Carrier Safety Regulations;" DOE 10 CFR

871, "Air Transportation of Plutonium," NRC 10 CFR 20-21-73; Interim Management Directive

5001; "Safety Health and Environmental Protection," of 9-29-77 to DOE 5481.1 "Safety,

Analysis and Review System," of 3-20-78.

Some of the key points of the regulations are presented here.

The type of packaging required in the DOE order is governed by the amount and radioactivity of

the fissile materials being transported. Fissile materials are defined as uranium-233 and 235,

plutonium-238, 239 and 241, neptunium-237 and curium-244. (It should be noted that neptunium

and curium are not listed as fissile in 49 CFR 173.403 and 10 CFR Part 71.)

The least restrictive package, is a Type A package. Type A packages cannot exceed a quantity of

aggregate radioactivity determined as A1 for special form radioactivity and A2 for normal form

radioactivity. Values for A1 or A2 quantities are shown in Appendix A, Table A-I, 10 CFR 71.

Special form radioactivity is radioactive material which is either a single solid piece or is contained

in a sealed capsule that can only be opened by destroying the capsule. The piece or capsule has at

least one dimension not less than 5 mm (0.197 inch) and it satisfies test requirements outlined in

10 CFR 71.75. The amount of radioactivity for the plutonium pits is too high to be shipped in

Type A package. Therefore, Type B packages are required. (They include the package and its

radi oacti ve contents.)

Another factor used to regulate shipments is the fissile classification. The material to be

transported is placed into one of three categories based on the controls needed to provide nuclear

criticality safety during transport. The categories include fissile class I which permits transport of

unlimited numbers of packages in any arrangement and requires no criticality control, and fissile

class II which permits the package to be transported with other packages in any arrangement

provided that they do not exceed an aggregated transport index of 50, with individual packages

having a transport index of not less than 0.1 and not more than 10. (The Index is a dimensionless

number rounded up to the first decimal place.) The transport index is further defined as a number

expressing the maximum radiation level in millirems/hour at 1 m from the external surface of the

package, or for fissile class II packages, the same definition for the number is obtained by dividing

4.4.2-2

the allowable number of packages which may be transported together, as determined in 10 CFR

71.59, by 50, whichever number is larger.

The final category is fissile class III which is a shipment of packages which is controlled in

transport by specific arrangements between the shipper and the carrier to provide nuclear

criticality safety. There are special requirements for plutonium shipments. The only requirement

applicable to plutonium metal is that plutonium in excess of 20 curies/package must be shipped

as a solid (which plutonium metal meets.)

The fissile class III shipments of plutonium would require a Type B package. These packages are

specified in 49 CFR 173.416. The package used is per DOT specification 6M which can be used

for only solid or gaseous radioactive materials that will not undergo pressure generating

decomposition at temperatures up to 121° C (2500F) and that do not generate more than 10 watts

of radioactive decay heat. A 6M package is a metal container that can contain no more than 4.5

kilograms of plutonium. This is due to the 10 watt decay heat limitation. It also provides a

double containment.

The Savannah River Site (SRS), for example, has four certified packages that are currently in use.

These are NRC approved packages (certification numbers 9965 through 9968.) These packages

are used for shipping solid metal or alloys of fissile or other radioactive material, as well as oxides,

scrap, or powders. The maximum payload of each of the four packages is 53 pounds. The gross

weight of the packages varies from 193 to 627 pounds. The difference in weight is due to

shielding requirements.

These SRS packages are placed in a rack, called a "bird cage," to maintain distance between

packages and placed on a special truck called an SST (Safe Secure Transporter.) A maximum of

125 of these packages can be shipped on the same SST.

Plutonium Oxide Transport

Plutonium oxide is not a nuclear weapon component and its transport is governed by DOE Order

5480.3, "Safety Requirements for the Packaging and transportation of Hazardous Materials,

4.4.2-3

Hazardous Substances and Hazardous Waste." The DOE Orderincorporates significant parts of

Title 10 CFR 71, "Packaging of Radioactive Material for Transport," which promulgates Federal

regulations for the packaging of radioactive materialfor transport and is the governing regulation

for civilian shipments of radioactivematerial.

Plutonium oxide must also be shipped in a Type B package. The same certifiedcontainers listed

above for plutonium metal shipments are also used to transport plutonium oxide. In addition,

there is another package, Certification #5320, that can be used for shipping plutonium and

americium oxides. This package uses a 10"by 12" diameteraluminum pipe as an inner container,

with a cylindrical stainless steel pressurevessel as the secondarycontainment. Thiscan is limited

to 357 grams of plutonium of any isotopic composition (403 grams of Pu0 powder) or 176 grams

of americium (200 grams Am0 powder.)

B. On-Site Shinments

A draft DOE Order 5480.x was issued in July 1991 to cover on-site "Packaging and

Transportation of Hazardous Materials Substances and Wastes." While not official, the Oak

Ridge Reservation (ORR) is following most aspects of the order. SRS has also adapted some of

the provisions. The order calls for site specific transportation safety manual to be prepared and

maintained for each site. The manual documents safety, health, and environmental protection

measures being taken for all on-site transfers of hazardous materials. While packaging

performance is the preferred way to ensure overall safety, DOE desires an integrated approach

which considers the packaging in combination with specified communication and controlmeasures.

The on-site packaging and transportation programs include the following elements: Identification

of responsibilities, lines of authority and program approval procedures, definition of safe

packaging requirements, descriptions of transportation systems, operational controls, safety

methodology, site descriptions, emergency response and process for non-routine packaging and

transportation activities. ORR has a Transportation Safety Manual outlining these activities and

SRS is studying issuing such a manual. At present SRS has a collection of these procedures. The

4.4.2-4

draft Order states that "Adherence to Federal standards, normally applicable to off site

transportation, is an acceptable approach to meeting on site standards."

Other factors also regulate on-site movement. The least restrictive is movement within a Material

Control Area (MCA). In this case there is less documentation and security measures required.

However, the same stringent safety and criticality controls are employed. In case of movements

between MCA's security surveillance is greater with constant surveillance of the transport'slocation and condition. If the routes selected for the movement can be cleared of other traffic, the

use of dedicated security and emergency response readiness can be reduced. If the transport is

across a public highway on the reservation, which is possible at many DOE sites, the shipment

would probably be treated as an off-site shipment and all the measures required for such

shipments must be followed.

4.4.2-5

4.4.3 TRANSPORTATION OF MOX FABRICATION FACILITY ANDREACTOR PLANT NUCLEAR WASTE

This section presents a study of the transportation of the waste forms from the ABWR and the

reactor complex fuel fabrication plant to on-site storage as well as transport to off-site DOE

disposal facilities. The study included the type of containers and shipping carriers required to

meet the applicable Federal, State and local regulations.

TRANSPORT OF FUEL FABRICATION PLANT NUCLEAR WASTES

Although a new fuel fabrication plant of modern design will be required, it is assumed that some

minimum amount of wastes would still be generated that cannot be recycled within the plant.

During the fabrication process, some of the material must be recycled because the do not meet the

final pellet specifications. This would include materials required for setting up machine

parameters, material for test specimens, material ground off during machining the pellets and

pellets with surfaces defects such as cracks. Either a dry recycle or wet recycle is used depending

on the purity of the material. If the material is out of specification because of damage, such as

chips or cracks, it is ball milled and the powder returned to the process head end. This is a dry

method. However, if it is powder from cleanup of the glove boxes which could contain

impurities, a wet method is used. This wet method consists of dissolving the material and

processing the solution by solvent extraction or ion exchange.

There are three categories of nuclear waste that can arise from fuel fabrication. These are solid

glove box waste, solid waste from retired equipment, and liquid waste from scrap recovery and

from work in the plant analytical laboratory. This waste can be classified as low level wastes,

mixed wastes (radioactive and hazardous waste), and transuranic (TRU) waste. Before the waste

can be disposed, it may be treated for: volume reduction of the waste; reduction of the plutonium

contained in the waste; and solidification of the liquid waste. A typical plant will usually use

compression or incineration to reduce the volume. Plutonium recovery is accomplished by ash

leaching, acid digestion and a washing process. Solidification of the caste can be by cementation or

bitmunization, such as is practiced in French and German plants. Some references state that less

than 0.5% of the throughput of the plutonium will be present in the waste and that more than

4,4,3-1

50% of this plutonium can be recovered. As an example, the new MELOX plant in France is

designed to recycle as much plutonium waste as possible. Low activity liquid wastes are treatedin the Marcoule treatment unit. At MELOX waste which can be burned will be incinerated with

the plutonium being recovered from the ash. The plant does not expect to produce more than 100

m3 of storable waste per year, of which only 4 m3 will require underground storage. MELOX is

a 120 MT/year plant.

Some DOE sites' have facilities to dissolve and recover the plutonium at the present time and

these could easily handle the additional feed stream from an on-site MOX plant. Where future

operations of these facilities at DOE sites are uncertain ,if these programs are not continued, a

plutonium disposition program might have to recreate what are already existing capabilities.

The majority of a fabrication plant waste can be expected to be the equivalent of Class A low level

waste with the small amount being equivalent to Class C low level waste requiting disposal in the

equivalent of Class B/C storage vaults. Any greater than Class C waste and TRU waste would

and could be disposed of at the WIPP or Yucca Mountain repositories. The disposition of mixed

wash from DOE facilities is not determined at this time.

A. On-Site Shipments:

The DOE Orders 5480.x apply to on-site waste shipments as well as to the on-site oxide or metal

shipments. At ORR for example, low level waste that can be compacted is placed in dumpster

type vehicles and driven to the waste burial ground where is compacted and placed into lined

trenches provided with a sealable cap. Non-compatible low-level waste materials are placed in 4'x

4'x 6' steel-boxes and transported to the burial ground. Material such shoe co,,ers, gloves, papers

are placed in stainless steel 55-gallon drums for storage. At SRS a Central Incineration Facility to

handle low-level combustible waste is planned but the project is currently on "Hold". At ORR,

an existing TSCA incinerator is currently accepting only low-level liquid waste, primarily of

contaminated oil. The oil is shipped in 55-gallon drums or in tanker trucks. While the facility

was designed to burn solids as well as liquids because of the large backlog of liquids there are no

existing plants to burn solids.

4.4.3-2

B. Off-Site Shipment:

DOE order 5480.3 covers off-site shipments of radioactive and hazardous waste. It is believed

that the vast majority of the waste generated by the fabrication plant could be retained on-site so

there would be no need to sl_ip off-slte. Where off-site shipment is desired however, a system

similar to that in use at ORR could be used. ORR has a contract with a private company which

can sort out clean material, and either incinerate or melt the remaining low level material. A

combination of drums and boxes, similar to what a private utility uses to handle shipments of low

level nuclear waste from reactors to storage sites, can also be used.

_TRANSPORT OF REACTOR COMPLEX NUCLEAR WASTES

Due to the dedicated waste volume reduction and plant worker radiation exposure reduction

features designed into the evolutionary ABWR, the type of waste from a mixed oxide fueled

ABWR will be similar in type but less in quantity than that generated in current BWRs. As a

benchmark for comparison of waste types, The River Bend unit of Gulf States' utilities is a 936

MWe BWR. This reactor is expected to generate 155 55-gallon drums (1,163 ft3)per year of dry

activated waste (DAW), Class A low-level waste, plus another 12 95 ft3 -boxes (1,140 ft3) of

the same waste. The dose rates on these waste range from 10 mRem/hour - 2 reins/hour. In

addition, about one 120 ft3 High Integrity Container (HIC) shipment/year are usually required for

resins which are a class B low-level waste with dose rates ranging from 8 Reins/hour - 30

Reins/hour. Riverbend also averages 45,207-ft 3 liners/year (9,315 ft 3) of Class A low-level

radwaste sludges. These sludges have dose rates from 0.5 Reins/hours- 10 Reins/hour. The

_proposed 1300+ MWe plutonium disposition ABWR has been designed to emphasize waste

management and would produce less waste than a conventional BWR. Reactors also generate a

limited amount of greater than Class C low level waste (GTCC) in the form of activated metals.

Radiation levels of this material can exceed 30,000 Rems/hour but only average one shipment/year.

This Type B material will also be generated when the reactor is decommissioned.

4,4.3-3

A. On-Site Shipments:

Again, the same DOE Orders regulate on-site reactor waste shipments as described above. On-

Site shipments are made using the 55-gallon drums and metal boxes for the DAW, Class A, low-

level waste as shipping containers. Shipments to storage areas is made by fiat bed trucks or

covered trucks. At ORR, for example, resins from the research reactors on-site are dried,

packaged in metal canisters, and buried. The Class A radwaste-sludge could be shipped in

disposal liners, as civilian reactors do; to ship these materials from the reactor to low level burial

grounds. The GTCC waste is normally shipped in spent fuel shipping casks which are heavily

shielded.

B. Off-Site Shipments:

Off-site reactor waste shipments will be regulated by DOE Order 5480.3 and as noted previously

use the same type of shipping containers and casks as civilian reactors.

4.4.3-4

4.4.4 Spent Fuel Transportation & Logistics

4.4.4.1 Summary

A review has been conducted to determine the current status of the Civilian

Radioactive Waste Management System (CRWMS). Using that information, the

potential impacts on spent nuclear fuel (SNF) discharged from the Advanced

Boiling Water Reactor (ABWR)in a weapons gradeplutonium destruction modeweredetermined.

Although a significant numberof studies andevaluations have been conducted over

the years related to permanent disposal of SNF and high-level waste (HLW), anumberof issues remain to be finalized. Most of the evaluations relatedtothis

program are in the conceptual design stage and are expected to be firmed-up over

the next several years. This includes, but is not limited to: site

characterization evaluations leading to final site selection for the Federal

Repository; repository spent nuclear fuel (SNF) storage capacity, repository

waste acceptance criteria; site selection and design of a Monitored Retrievable

Storage (MRS) facility; design of the Multi-Purpose Canisters (MPC); and, design

of transportation casks for SNFacceptance, handling, transportation, storage and

disposal.

As a result of this review,no significantconcernswere identifiedrelated to

acceptance,handling,transportation,and disposal of spent mixed-oxide (MOX)

fuel from an ABWR in a plutoniumdispositionmode. Likewise,few firm design

requirementswere found which could be used at this time to firm up the ABWR MOX

design and other ABWR designor processissuesto ensure compatibilitywith plans

for the CRWMS.

Compatibilitywith the CRWMS is highlydesirablefor economicand other reasons.

To achieve this, it is recommendedthat an on-going awarenessof the status of

the CRWMS be maintainedfor potentialimpact. The Departmentof Energy (DOE) is

strongly encouraging continuing involvement and input from stakeholdersto

influencedesign and implementationof the CRWMS.

4.4.4-1

I!

4.4.4.2 Discussion

The purpose of this task is to: 1) present the results of a review of the current

plans to ship SNFfrom civilian nuclear power plants to the Federal Repository;

and, 2)assess the range'of options available for spent nuclear fuelfroman ABWR

tn a plutonium disposition modeto meetthe projectneeds. A,significant amount

of the information presented below is still conceptual in nature. It is expected

that studies by DOE and support_contractors over the next several years will

• result in firming up information of relevance to destruction of weapons grade

plutonium in the ABWR.

A comprehensivereview of the entire CRWMS is currentlybeing conductedby the

Secretaryof Energy. Resultsof this review is scheduledto be completedin 30-

60 days and may presentadditionalpolicydirectionof relevanceto this program.

DOE StandardContract

A standardcontract for transportof SNF and high level radioactive(HLW)waste

is required by Section 302 of the Nuclear Waste Policy Act of 1982. This

contract iscontainedin 10CFR g6 which includescontractualprovisionsrelating

to disposalof SNF and HLW from civiliannuclearpower reactors requiredto be

licensed under Sections 103 or I04(b)of the 1954 Atomic Energy Act as amended

(42 U.S.C. 2133, 2134(b)).

Under provisions of this contract, the DOE will make available nuclear waste

disposal servicesto the owners and generatorsof SNF and HLW. In exchangefor

these services,the ownersor generatorsof such fuel or waste pay fees on a full

cost recoverybasis into a NuclearWaste Fund. Thesecontractshad to be entered

into by June 30, 1983 or by the time generationof spent fuel or high level waste

generation is initiated,whichever is later. These contracts provide for: I)

title transfer of SNF and HLW to the DOE at the utilitiessite; 2) shipmentof

the SNF and HLW to the FederalRepository;and 3) disposalof the SNF and HLW by

the DOE followingcommencementof the operationof the repository.

4.4.4-2

It is assumedthat the contractual provisions of 10 CFR961 would be waived for

the spent ABWRHOX fuel since it is owned by DOE. However, the technical,

financial and programmatic aspects of the Standard Contract are likely to apply.

DOEwtll provide a cask for shipment of the SNF and/or HLWfrom the utilities

nuclear power reactor, or such other locat|on designated by the user, to the DOE

st(}rage facility(s). The cask will be delivered to the user sufficiently in

advance of the shipment, suitable for use at the site, meet applicable regulatory

requirements and be accompanied by the following information: 1) written

procedures for cask handling and loading; 2) specifications for user furnished

canisters for containment of failed fuel; 3) training for user personnel in cask

handling and loading; 4) technical information, special tools, equipment, lifting

trunnions, spare parts and consumables needed to perform incidental maintenance

on the cask; and 5) documentation on the equipment supplied by DOE.

The user will be responsible for incidental maintenance, protection and

preservation of the casks provided to the user. The user will also be

responsible for providing all preparation, packaging, required inspections and

loading activities in preparation for transportation of the SNF and HLWto the

DOEstorage fac i 1i ty (s).

A prioritywill apply for shipmentof SNF from generators. This priorityranking

is issuedby DOE in aF annualAcceptancePriorityRanking (APR) reportwhich is

based on the date the SNF was dischargedfrom the reactor. The oldest fuel or

waste on an industry-widebasis will have highest priorityfor shipmentto the

repository. Reinserted SNF will be removed from the APR and rescheduledfor

shipmentin accordancewith a new APR based upon it's permanentdischargedate.

DOEhas indicatedthat it will accept "standardnuclearfuel" asdefined by fuel

specificationsin the contract. These specifications,inessence,establishthe

waste acceptance criteria of spent nuclear fuel for disposal in the Federal

Repository. DOE will evaluate the feasibility of disposing of non-standard

nuclear fuel and inform the utility of any adjustmentsthat may be required.

Standard nuclear fuel specificationsinclude:

4.4.4-3

• Maximumphysical dimensions)

Maximumnominal physical dimensions are as follows:

Overall length: BWR--14 feet, ]1 inches; PWR--14 feet, ]0 inches

" Active"fuel lenqth: BWR--12 feet, 6 inches; PWR--12.feet, O.inches ....,_

_ross sectlon: BWR--6 tnches x 6 inches (not Including -the channel);.,,PWR--9inches x 9 inches

• Non-fuelcomponents

Non-fuel components including, but not limited to, control spiders, burnable

poison rod assemblies, control rod elements, thimble plugs, fission chambers, and

primary and secondary neutron sources, that are contained within the fuel

assembly, or BWRchannels that are an integral part of the fuel assembly, which

" do not require special handling, may be included as part of the spent nuclear

fuel delivery for disposal.

• Minimum cooling time

The minimum cooling time for fuel is five years.

• Non-LWR fuel

Fuel from other than LWR power facilitiesshall be classifiedas non-standard

fuel. Such fuelmay be unique and requirespecialhandling,storageand disposal

facilities.

• Consolidatedfuel rods

Fuel which has beendisassembled and storedwiththe fuelrods in a consolidated

manner shall be classifiedas non-standardfuel.

• Failed fuel

Assembliesshall be visuallyinspectedand those which are structurallydeformed

or have damaged claddingto the extent that specialhandingmay be required,or

4.4.4-4

for any reason cannot be handled with normal fuel handling equipment, shall beclassified as failed fuel.

A revision to the current Standard Contract is expected to be developed and

• published inthe'Federal Register by the DOEfor commentduring CY-Ig94. This

revision is expected to clartfy, modify and add_several_newprovisions to the

contract, including changesto someof the standard nuclear fuel specifications.

This is expected to include a possible modification of the spent nuclear fuel

cooling time, establish criteria for and disposition of failed fuel, address

priority for disposing of SNF from permanently shutdown reactors, address

disposition of consolidated fuel rods, and other potential changes.

SNF Canisters

Studieswere initiatedin 1992 to determinethe feasibilityof using sealed

canisterstoaccommodateSNFduringwasteacceptance,transportation,storageand

disposaloperationsthroughoutthe CRWMS. Initiationof thesestudiesresulted

primarilyfrom concernsof repeatedSNF assemblyhandling and packaging

operationswhichwouldbe requiredwith otheroptions.

DOE iscurrentlyevaluatinga September30, 1993draftof the MPC Implementation

ProgramConceptualDesign Phase Report. Productsfrom this study include

conceptualdesigns for the MPC, the transportationcask, the monitored

retrievablestorage(MRS)facilityand the utilitytransfersystem. The MPC

conceptualdesign includes canister configurationsfor containingboth

pressurizedwaterreactor(PWR)and boilingwaterreactor(BWR)SNF assemblies

in the same MPC. The transportationcask conceptualdesignsprovidecasks

necessaryfor transportingMPCs fromwastegeneratingand storagesitesto the

repository. TheMRS facilityconceptualdesignprovidesfacilitiesfor loading

MPCs and storingthem untilthe permanentrepositorybecomesoperational.The

utilitytransfersystemconceptualdesignprovidesan on-sitetransfersystemat

commercialreactorfacilitiesto load,handleand storeMPCs.

4.4.4-5

The MPCpreliminary design concepts are based on several assumptions:

• A legal weight truck caskwith a capacity of 4 PWRor g BWRassemblies, nominal

2S-ton loaded weight, for facilities with limited cask handling capabilities.

• A mediumsize MPCfor rail transportation with a capacity of 12_PWRor_24 BWR

assemblies (without reliance on burnupcredit for either type of fuel). This

will involve a nominal 75 ton loaded weight MPCin a transportation cask.

• A largesize MPC for rail transportationwith a capacityof 21 PWR (with

relianceon burnupcredit)or 40 BWR assemblies(withoutrelianceon burnup

credit). This will involvea nominal 125 ton loaded weight HPC in a

transportationcask.

• SNFwould be initiallyacceptedat theMRS in theyear2000. Betweentheyear

2000and 2010all SNFwouldbe transportedto the MRS for temporarystorage.

• Prior to 2010, all MPCs receivedat the MRS would be removedfrom the

transportationcask and placedin a steelor concretestoragecask for storage

at the MRS. Any SNF assembliesreceivedin legalweighttruckcaskswouldbe

unloadedat the MRS and placedin eitherof the two MPCs and then placedinto

storagecaskspendingsubsequentshipmentto the repository.

• Beginningin2010,theMRSwouldserveas a stagingarea. MPCsarrivingat the

MRS wouldbe placedon dedicatedtrainsfor shipmentto the federalrepository.

The MPCs consistof a cylindricalshellwithtwo lids,a spentfuelbasketand

a shieldplug. ThespentfuelbasketprovidesstructuralsupportfortheSNFand

a mechanismforthetransferof theheatgeneratedby theSNFintotheMPCshell.

Thespentfuelbasketalsoprovidescriticalitycontrolto ensuretheSNFremains

subcriticalunderall definedcircumstances.

The largeMPCdesignrequiresburnupcreditfor shipmentof PWRSNF. Themedium

12 PWR basketdesignemploysa water gap flux trap arrangementto improve

effectivenessof the boratedaluminumneutronabsorberpanels. Similarlythe

4.4.4-6

smaller size of a BWRfuel assembly allows the borated aluminum panels to come

closer to the center of each fuel assembly which results in the neutron absorber

panels being more efficient. Using this design approach, the BWRbasket

configurations do not require burnup credit for storage or transportation

licensing with either of the two MPCs. Criticality requirements wtll need to be

calculated for spent ABWRfuel to ensure ,compliance,with the preliminary MPC

designs.

A SNF cladding temperature limit of 340 C was adopted as the design limit for

preliminary MPCthermal analyses to ensure that storage limits could be met in

the transportation cask design. This is the 10 CFR72 storage temperature limit

for ten-year cooled fuel. ]0 CFR 73.60 and 67 state that the radiation dose

after a two year cool down period must be at least 100 rem/hr at a distance of

three feet from all surfaces of the fuel bundle. This is to provide protection

_against theft or diversion. _ Structural design of the MPCwas based on a 60 g

acceleration limit for the hypothetical 9-meter drop accident scenario. This

represents the maximumacceleration that light water reactor fuel assemblies must

structurally withstand in a side drop without failing. Calculations of the

cladding temperature and radiation dose for spent ABWRfuel under the above

conditions and structural integrity for the various accidents involving handling

and shipping of ABWRspent nuclear fuel need to be determined to ensure

compliance with these preliminary requirements.

Someconsideration is being given by DOEto design a smaller MPCfor legal weight

truck use. Currently, there are about nineteen reactor sites that are limited

to legal weight truck cask shipments. Most reactor sites will be able to use the

125 ton MPC.

SNF T.ransportationCasks

The Office of CivilianRadioactiveWaste Management (OCRWM)has an objectiveof

developing and placing into operationa system capable of transportingspent

nuclear fuel and high-level waste from the various waste sources to waste

receivingfacilitiesbeginningin January,2000 and subsequentlyto the Federal

Repository.

4.4.4-7

The Department of Energy has sponsored a number of transportation cask design

efforts to accommodatethe various assemblies expected to be accepted for

disposal. The MPC is sealed and placed inside the transportation cask for

shipment by rail to the MRSor Federal Repository. At this time, there are

conceptual designs for two transportation casks. One is to transport the large

125 ton MPC and the second is to transports,the 75 ton MPC.-These_wetghts

represent the under the hook weight at the reactor spent fuel storage pool and

includes the MPC (with outer lid removed), SNF, water, transportation cask body

and lifting yoke.

Design temperature limits and acceleration values for preliminary MPC

transportation cask analysis and design are the sameas used for MPCanalysis and

design (i.e. 340 C and 60 g acceleration). The cask systems must be certified

by the Nuclear Regulatory Commission (NRC) and must comply with limits imposed

by various'agencies including the DOE, NRCand the Department of Transportation

(DOT).

MRS Storageof SNF

Without the availabilityof a Federal Repository, inventories of SNF have

continuedto build in the storagepools at reactorsites,forcingmany utilities i

to take actionsto increasethe SNF storagecapacitiesat their sites. The most

prevalent action taken by the utilitieshas been to re-rack their spent fuel

storagepools with highercapacitystorageracks. A few utilitieshave used fuel

assemblyconsolidationas a means of increasingthe capacityof their pools,and

a few more have installeddry storagefacilitiesof varioustypes on their sites

to permit storageoutsideof their reactorstoragepools. Currently,fiveonsite

spent-fueldry cask storagefacilitieshave site-specificlicensesfrom the NRC

and are in operation.

The DOE was authorizedby Congressin the NuclearWaste PolicyAmendmentsAct of

1987 to develop a Monitored Retrievable Storage (MRS) facility to provide

temporary above-ground storage for a limited amount of SNF from commercial

reactors. For planningpurposes,the MRS iscurrentlyscheduledto receivespent

nuclearfuel from utilitiesin January, 2000. A recommendationfor locationof

4.4.4-8

the site of the MRSis currently scheduled for September1994 with start of

construction in September1998.

The MRSfactllty conceptual design provides facilities for temporarily storing

MPCscontaining SNFunttl the permanentFederal Repository becomesoperational.

MPCsarriving at the HRSare removedfrom the transportation caskrplaced wtthtn

a disposal container designed for permanentdtsposa] and then placed within a

vertical steel or concrete structure for temporary storage at the MRS. The MRS

conceptual design also allows handling of bare SNFassemblies.

A dedicated factltty for inspecting, testing, andmaintaining the cask systems

was recommendedby the General AccountingOffice as the best meansof assuring

their operational effectiveness, safety, and regulatory compliance. In 1987,

OCRWMrequested a feasibility study be madeof a CaskMaintenanceFacility (CMF)

that would perform the required functions. It is currently envisioned that the

CMF will be integratedwiththe MRS facility.

With regardto the ABWR,itwouldappearthatmaximizingthe spentfuelstorage

capacityin the reactorbuildingwouldbe prudentbecauseit wouldminimizethe

handlingnecessaryto placespentfuel back intothe reactor,if desired,for

additionalpowerproduction.This wouldalso appearto be cost effectiveas

comparedto constructionof a centralizedstoragefacility. In any event,a

minimumtenyear storagecapacityat the ABWRsitewouldappearto be necessary

beforeshipmentto theMRSor repositorywouldbe allowedbecauseof a numberof

factors,including:I) existenceof the currentlargebacklogof spentnuclear

fuel;2) DOE'srequirementto givepriorityforshipmentof theoldestspentfuel

to the repository;3) possiblehigh priorityto be given to utilitiesfor

disposingof SNFfrompermanentlyshutdownreactors;and,4) sufficientcooldown

to meet the maximumcladdingtemperatureof 340 C.

Currently,defenseSNF is precludedfrom being storedin the MRS. It is

expected,however,thatthisprohibitioncouldbe eliminatedas discussedbelow.

4.4.4-9

Federal Repository Storaae

Site characterizationof YuccaMountain,Nevada is currentlyin progressto

determineit'ssuitabilityfordevelopmentasa FederalRepository.The Federal

Repositoryconceptualdesignwillaccommodate70,O00MTUof waste:63,020MTU of

SNF; 6,340MTU of DefenseHlgh-LevelWaste_(bHLW)and.640MTU of West-Valley

High-LevelWaste(WVHLW).Theearliestcurrentscheduleforcomenclngoperation

of the FederalRepositoryis 2010.i

The repositoryisbeingcreatedundertheNuclearWastePolicyAct (NWPA)of 1982

(U.S.C.Title42 I0101,as amended)andDOE was designatedas the ownerof the

facility. The law initiallyprecludedwastegeneratedas a resultof defense

activities.However,in 1985PresidentReagansignedan authorizationto allow

acceptanceof defenseHLW into the facility. Scopinghearingsare now in

progress to allow DOEto accept defense SNFassemblies into the repository.

DiscussionswithDOEstaffwereinitiatedto determineiftherepositoryisfully

committed.Thisinformationiscurrentlynotavailablebecausetheabovestorage

capacityis stillconceptual. By the year 2000, about40,000MTU of spent

nuclearfuel will existat nuclearreactorsites. By the time the last NRC

licensefor the currentgenerationof nuclearreactorsexpires,an estimated

totalof 87,000MTU willhavebeengenerated.Ifthe YuccaMountainrepository

is limitedto 63,020MTU of SNF,additionalpermanentdisposalcapacitywillbe

required. SNF for permanentdisposalfrom the ABWR may be placedin either

facility.

The waste package design criteria for the Federal Repositoryhave been

incorporatedinto the MPC conceptualdesign in order to meet repository

conceptualrequirements. A numberof importantdesign requirementsof the

repository,however,will not be finalizedfor severalyears. One of these

includesthe thermalloadingstrategyto supportthe licensingprocess. This

issueinvolvesthe thermalloadingof the repositoryin the near-fieldandfar-

fielddue to heat generatedby the emplacedwaste. Becauseof the schedule

associatedwith thermalloadingstudies,designof the MPC is expectedto be

finalizedbeforeit is knowniftheMPCdesignis compatiblewiththe repository

4.4.4-10

design. It may be necessary to repackage the SNFat the MRSor at the repository

in order to meet the thermal loadtng criteria. Alternatively, the MPCdesign

could be subsequently modified to be compatible with repository requirements.

MOXfuel wtll have a different heat load than U02 fuel. Decay heat calculations

as a function of time will need tobe performed for the-ABWRfuel as comparedto

BWRfuel assemblies from commercial nuclear plants.

A number of interfaces exist between the MPC and the repository facilities.

These include the possibility of a disposal overpack, waste handling building,

waste package transporter, and the subsurface emplacement operations. Each of

these interfaces were evaluated and considered in the MPCconceptual design.

In federal laws, regulations, and departmental directives, the US Congress, the

r Nuclear Regulatory Commission, and the US Department of Energy have developed

criteria for retrievability of waste emplaced in a geologic repository for high-

level radioactive waste. In response to these criteria, the Yucca Mountatn

Project is expected to have a capability to retrieve emplaced waste as a planned

contingency operation.

10 CFR 60 requires that "... a nuclear criticality accident is not possible

unless at least two unlikely, independent, and concurrent or sequential changes

have occurred in the conditions essential to nuclear criticality safety." One

of these changes must be the addition of amoderator- water. Concern is limited

to accident scenarios involving flooding of the repository.

During repository operation, bare fuel assemblies would be vulnerable to flooding

only in the hot cells of the waste handling building. Careful facility design

is expected to reduce the risk of a criticality accident during fuel handling

operations virtually to zero. In a11 other near-term repository environments,

the fuel is protected by watertight casks or disposal containers. Under post-

closure conditions involving intact spent nuclear fuel assemblies, it is unlikely

that criticality will be found to be a problem for any container configuration.

4.4.4-11

This wtll need to be vertfled for ABWRMOXfuel as compared to commercial BWR

fuel assembltes.

In the longer term (hundreds to thousands of years after repository closure),

there is a potential concern for container failure and flooding. In the very

long term, the 'containers and the_fuel..assembltes _themselves may_.have

disintegrated; tnthat case, the physical configuration of the fuel at the ttme

of emplacement is more or less academic. Current plans for analyzing post-

closure'criticality scenarios are based on the assumption that water Intrusion

has occurred. Criticality concerns with ABWRMOXfuel wt11 need to be determined

as compared to commercial BWRfuel assemblies for these conditions.

4.4.4.3 RecommendedFollow Up Actions

The most cost effective-approach for spent fuel managementwould be to ensure

compatibility of the spent ABWRMOXfuel assemblies with the CRWMS. To achieve

this goal the following actions are recommended:

• Maintain cognizance of the status of the Civilian Radioactive Waste Management

System (CRWMS). One of the first design finalization activities involves

issuance of an RFP for design of the MPCin the Spring of 1994 and award of the

contract in late 1994 or early 1995. Compatibility of ABWRMOXfuel with the MPC

design(s) would be critical.

• Defense SNF is currently not allowedwithin the MRS or Federal Repository.

Actions are being initiatedby DOE to allow defense SNF into both facilities.

The status of these initiativesneeds to be monitored and follow up actions

taken as necessary.

• Modificationsto the StandardContractare expectedto be publishedfor comment

during CY-1994. The modifiedcontractshould be reviewedand commentsprovided

to ensure that the StandardContractadequatelyenvelopesdesigncharacteristics

of spent ABWR MOX fuel.

4.4.4-12

• The MPCsystem (canister and transportation cask) must be licensed by the NRC

under a numberof regulations, including, but not ]imtted to: 10 CFR50, Domestic

Licensing of Production andUtilization Facilities; 10 CFR60, Disposal of High-

Level Radioactive Wastes tn Geologic Repositories; 10 CFR71, Packaging and

Transportation of Radioactive Material; and, IOCFR 72, Licensing Requirements

for the Independent Storage of Spent.Nuclear,.Fuel.-and_High-LevelRadioactiveWaste.

An effectiveinterface:should be established with the DOE Contractor(s)

developing the safety, environmental and other documentationfor this licensing

activity. Information related to the ABWRHOXfuel mayhave to be provided to

the Contractor(s), in addition to the specific issues identified below. This

will ensure that issues pertaining to spent ABWRMOX fuel are adequately

enveloped by the licensing documentation.

• Criticalitycalculationsneed to be performedfor spentABWR fuelto ensure

criticalityrequirementsaremet forthe MPCdesigns,transportationto theMRS

and repository,and storagewithinthe repository.

• Claddingtemperaturesof the ABWR MOX spent fuel assembliesneed to be

determinedto ensuremaximumtemperaturesdo notexceedthe MPC designlimitof

340 C followinga ten-yearcooldown.

• Structuraldesigncapabilitiesof the ABWRMOX spentnuclearfuelwithinthe

MPC need to be determinedto ensurestructuralintegrityduringhandlingand

transportationaccidents.

4.4.4.4 References

I0 CFR Part 961, StandardContractfor Disposalof SpentNuclearFuel and/or

High-LevelRadioactiveWaste

SAND86-2357,OGR Repository-SpecificRod ConsolidationStudy"Effecton Costs,

Schedules,andOperationsat the YuccaMountainRepository,December,1988

4.4.4-13

SAND89-7009,AlternativeConfigurationsfor the Waste-HandlingBuilding at the

Yucca Mountain Repository,August, 1990

ORNL-TM-II019,FeasibilityStudy for a TransportationOperations System Cask

MaintenanceFacility,January, 1991

PNL-SA-19567,US Programfor Managementof Spent Nuclear Fuel, April, 1991

............_'DOE/RW-O31IP,'_.A;,Monitored,RetrievableStorage.Facility:Technical Background

Information,July, 1991

SANDB7-2777,RetrievalStrategyReport for a PotentialHigh-LevelNuclear Waste

Repository.Yucca Mountain Site CharacterizationProject,December, 1991

" " PNL-8072,ForeignExperienceon Effectsof ExtendedDry Storageon the Integrity

of Spent Nuclear Fuel, April, 1992

DOE/RW-0419,1992 Acceptance PriorityRanking,May, 1992

DOE/RW-0407,Designingthe MRS, March, 1993

DOE/RW-0422, FY 1992 Annual Report To Congress,Office of Civilian Radioactive

Waste Management,July, 1993

OCRWM Bulletin,MPC ConceptualDesignReport Submittedto OCRWM, November,1993

4.4.4-14

4.4.5 COMPARISON OF U.S. AND :,NTE_'.NATIONAL

TRANSPORT REGULATIONS

No clear regulatory orenvironmental framework :exists-in .the-,US.for the ..use of recovered

plutoniumin MOX fuels, thereforetherehas been no developmentof a commercialinfrastructure

to support transport, safeguards or security for such transports. The "system" for the movement

•.........:"ofsuch materials isoperated-bythe.US:Departmentof Energy (DOE) and.the..US .Departmentof

Defense (DOD). The "system"was designed to support the transportof plutonium for fabrication

into nuclear warheads and the transport and deploymentof assemblednuclearwarheads.

The development of a commercial system will require a stated US policy for the civil use of

plutonium in MOX fuels and the development and completion of an Environmental ImpactStatement to reflect this action.

Although transportof radioactivematerialdates back to the beginningof the nuclearindustrythe

rapid development of nuclear plants and international trade in fuel cycle services such as

enrichment,reprocessingetc., have led to the evolution of an international transport infrastructure

to service the industry. In particular there is experience outside of the US of international

commercial transport both of plutonium and MOX fuel assemblies under the IAEA regulatoryframework.

Advances in package design and technology have been led by increasing emphasis on safety

assuranceand compliancewith transport regulationswhichin manycases exceed those appliedto

other dangerous goods. In the case of certain materials, security during transport has equal

emphasiswith safety in order to prevent theft or diversion of the cargo. Such security poses

special problems, and can only be described in general terms for obvious reasons.

4.4.5-1

4.4.5.1 National and International Regulations

A TransportRegulations

A review of the regulationsrelatingto plutoniumtransporthas identifiedthe following which are

mostapplicable:

National (US)a) 49CFR Pt 170 to 178 - Transportation

b) 10CFRPt 71 - Packaging of radioactivematerialfortransport and transportationofradioactivematerialundercertain conditions.

c) NUREG 0360 - Qualificationcriteria to certify a package for air transportation of

plutonium

d) US Department of Transport Specification 6M International

e) International MaritimeDangerous Goods (IMDG) Code.

f) InternationalCivil AviationOrganization(ICAO).

g) International AtomicEnergyAgency (IAEA) Safety Series.

No. 6 - Regulations forthe Safe Transportof RadioactiveMaterial.

A study of regulations a, b, e, f and g shows that these are essentially consistent in terms of

package, labelingetc. requirements,because they reflect or cross referenceto IAEA Safety series

No. 6. The basic principle of Safety Series No. 6 is that the safety of the package is vested in its

design and is therefore independent of transport mode i.e., land, sea or air. However within the

US the adoption of the mode specific NUREG 0360 together with the Murkowski amendment

places more test requirements on packages designed for air transport of plutonium. US DOT

Specification 6M is limited to packages containing very small quantities of material (less than

10Wheat output).

B. IAEA Regulations

It is important to note that the arrangementsfor transportationof certain plutonium by BNFL are

in accordance with the IAEA regulations to which the US is also a signatory. The transport of

irradiated fuel from Japan to Europe by the BNFL subsidiaryPacific Nuclear Transport LTD

(PNTL) is also carriedout underarrangements which comply with IAEA regulations.

As already stated according to IAEA philosophy the safety of the consignment is vested in the

packaging. As such they are subjected at all levels to stringent controls on design, manufacture

and operation.

4.4.5-2

Packaging varies according to the nature of the material carried and its radiological

characteristics. Complex methodologies involving dose uptake through direct or indirect

radiationcontact, pathways to the environmentetc., have been derived to classifyeach radioactive

elementor combinationof elementsaccording to hazard so as to permitthe selection of the most

appropriatetype of packaging.

In general, like the products of reprocessing, plutonium requires packages which have to

demonstratethat they retain their integrity in the most severe accident condition as defined by a

'..series of sequential tests: The so called IAEA type (B) tests involve a rigorous regime of impact

onto an unyielding target followed by an all engulfing fire to prove that the containment system

remains leak tight to the prescribedlimitsand that radiationlevels from the damaged package do

not pose an unacceptable threat to the public following such an accident.

........... Plutonium transportpackagingcalls for diversepackagingtypes to caterfor its many forms, from

powder to complete MOX fuel assemblies for fast and thermal reactors; each must serve the

conditions of safety and protection of the public. Although in very different forms, the

containmentstandardof packaging is the same.

The requirements of the IAEA Regulations, with regard to the standards of leak tightness, are

verydifficult for plutonium. With the mixtureof isotopes derived fromreprocessing of LWR fuel

the regulations restrictthe allowable leakage of plutonium dioxide to approximately 0.004 mg/h

under normal conditions of transport,andto 3 mg per week under accident conditions.

C. Package Approvals

Although this overlaps the section on package availability,it is appropriate to consider the topic at

this stage as it is clearly influencedby regulatory matters. Currently non-US packages would not

necessarily be approved by a US competent authority. However, 49CFR does allow for foreign

approved packages to be re-validated for use in the US by the appropriate US competent

authority. It also specifies that an application for re-validation must be made at least 45 days prior

to the required date for use of the package. There are no foreign packages presently qualified to

NUREG 0360 and therefore the air transport of plutonium is not seen as being applicable to this

project. The Murkowski amendment to NUREG 0360 effectively stops air transport in the US.

Consequently, only land and sea options are being considered.

4.4.5-3

4.4.5.2 Packages to Transport Plutonium or Plutonium Bearing Materials

A review of US packages has been carded out. It is doubtful if any of these could be used for

international transport as none are validated by IAEA Competent Authorities. There are no

Safkeg (2816C-16 Kg) type packages available and following the Murkowski Amendment to

" " NUREG 0360 (21.12.87), which for all practical.purposes.precluded.air,transport;of plutonium,

no developmentprogramshave been pursued.. Similarly,there are no packages in the scope of the

BNFL 1680 package.

....Tostart a plutonia campaign, the existing European packages could formthe basis for a transport

program. However, none of these are approvedfor use in the US and the US licensing process is

expensive and time consuming. Furthermore,thereare some significantdesign requirementse.g.,

double containment boundary, which have to be proven to meet licensing requirements. (It is

....... worth noting that.a Croft,AssociatesSafkeg was once validated for US use.)

Clearly for the movement of plutonia in any significant quantities a development program will

have to be started to design, license and fabricate a new generation of plutonium transport

packages. These will be doubly contained packages licensed by USNRC. They will undoubtedly

be truck or railtransported to the nearest ocean shipping port.

A search of European packages to answer an inquiry for BWR Fuel in general would not

necessarilycover the particularfuel design required.

Without more detailsof the reactorwe can only list packages for BWR Fuels of varying designs.

Firstly, there is a list of powder and metal transport packages alreadyin use for domestic and/or

internationaltransport, followed by severalfuel assemblypackages.

4.4.5-4

CONTENTS PACKAGE MAX CONTENTS(SUBJECT TO CONDITIONS)

PuO2POWDER GB/2816c/B(U)F 18 Kg MAX. *GB/3405A/B(U)F 2.5 Kg MAX. *GB/1680/B(U)F 72 Kg MAX.//

TNB/0145/B(U)F 1.5 Kg *

MOX POWDER, } GB/2816E/B(U)F 10 Kg MAX. *PELLETS }AND SMALL RODS } TNB/OI45/B(U)F 4.5Kg MAX. *PLUTONIUM METAL GB/2816E/B(U)F 4.6 Kg MAX. *

GB/3405A/B(U)F 4.5 Kg MAX. *BWR ASSEMBLIES TN 17 8 ASSEMBLIES /

EX ,4 12 ASSEMBLIES /FS 74 4 ASSEMBLIES//

• Approved to IAEA Regulations generally for all modes of Transport// Under application for IAEA Regulations/ In development stage for fresh MOX fuels. Already approved for irradiated fuel assemblies.

For PWR assembly transport the TN/0176 and COGEMA FS 69 are also used for international

transport in Europe (2 assemblies per package).

4.4.5-5

4.4.5.3 Vehicles

A. Air Transport

AirTransport- US

• As statedearlierthe Murkowskiamendmentto NUREG 0360 effectivelystops theair transportof

plutoniumin the US althoughairtransportis used effectivelyin Europe.

Air Transportof Plutonium- Europe

Transport between countrieson MainlandEurope can generally be accomplishedby road alone

but for particularly long shipments and for current shipments to and from the United Kingdom

(includingsome UK domesticshipments), airtransport is efficientlyemployed. By this means,the

timeto accomplish an internationaltransport is reducedto a few hours, during which time the

materialis removed entirelyfrom the public view with consequent security benefits. Although it

is not necessary to obtain Competent Authority package approval for countries which are

overflownby suchtransports, it is necessaryto obtaina non-scheduledflight clearancefrom thosecountriesaviationauthorities.

B. Land Transport

Road-US.

Our currentknowledge indicates that the entire production and fabrication of plutonium has been

controlledby the US DOE. Similarlythe transport of plutonium has been controlled by the US

DOE Transportation SafeguardsDivision (TSD) based in Albuquerque.

This group has been responsiblefor the transport fleet, the physicalprotection measures built into

the fleet and the armed escorts that accompany each shipment. They also provide the

management,schedulingof vehicles (tanks and trailers) trainingand maintenance. There are no

commercial alternativesinvolvedin plutoniumtransport.

We understand that Tri State Motor Transport (TSMT) used to provide high security services

with armed escorts and armored vehicles. Changes in USNRC policy with respect to safeguards

and a decliningmarket has led to TSMT discontinuing this service.

Currently TSD is providing weapon returns transport for the USDOE. If this work load declines

they may be able to provide a safeguarded transport service under appropriate sub-contract

4.4.5-6

arrangements.

Road TransportExperience- Europe.

Plutonium in its many forms attractsthe most difficultsecurity requirementsduringtransport. In

Europe all plutonium materials from powderto fuel assemblies in Category I quantities (2 kg or

more) are transported in more or less the same way, since the regulations do not currently

recognize the differencein form in which plutoniummaterialsexist; (for example MOX could be

considered as possessing an intrinsic,security characteristic resulting from .the dilution .of

plutonium by up to 20 times ina uranium matrix). Vehiclesused to transport these materialsby

road are specially constructedaccording to national standards, to provide an effective barrierto

attempts to penetration by an adversary,or by special devices to resist theft of the vehicle itself,

details of which cannot be given here. In general these vehicles present, as far as possible, the

.............same appearance as other haulage vehiclesof similartype. Irrespective of any escorting forces the

vehicles themselves are in constant communication with their operations control center and in

most cases they are tracked by automatic systems to give a. constant indication of their position

and status during transport. Special package tie-down systems are included in the vehicle

construction. Where "standard"packages are concerned it is possible to consider an integratedhandling system which is able to load or unload vehicles in despatch on receipt facilities with

minimumcloseuptake to transportworkers. During road transportsthe vehicles are accompanied

by armed escorts whose purpose is to further enhance the security of the shipment as well as

being an additional communication channel to the operations center. In some cases this escort, is

provided by a special constabulary which is specifically empowered by law to protect such

transports. They may also, in addition, be accompanied by the civil police force who can

generallyperform extra duties such as smoothing the traffic flow to allow the unhinderedtransit

of the security vehicle and its escort.

Rail

While rail transportation does provide benefits in terms of carrying very heavy loads over long

distances it also has its own unique safeguards.problems e.g., evasion of a potential threat is

difficult. It has not been possible to pursue this any further within the scope of this study.

Rail transport of Category I materials outside licensed sites is technically feasible under the

guidelines provided in INFCIRC/225 and is covered by legislation in European nations, however

it is not known to what extent actual transports are undertaken using this mode. The use of

through rail transport, like sea and air, will usually involve a road movement either at one or both

4.4.5-7

ends since many facilities do not possess a rail head or sea terminal or airport within the site

boundary and so road transport can never completely be replaced.

Sea Transport

BNFL Transport Division has many years experience of transporting .nuclear.materials by.sea

using PNTL's purpose built fleet of vessels. BNFL Transport Division.vessels.have completed

over 130 voyages and covered around 4,000,000 nautical miles.

" - The latest revision ofthe International Maritime Organization (IMO) regulations contain limits on

the quantities of, for example, MOX fuel, which can be carried by certain types of vessel. The

IMO formulated a code for transporting irradiated nuclear fuel, plutonium and high level

radioactive wastes in October 1993. The requirements of the code will be promulgated tomember states in 1994.

The code applies to new and existing ships regardless of size including cargo ships of less than

500 tons gross tonnage, engaged in the carriage of irradiated nuclear fuel, plutonium and high

level _'adioactivewastes in flasks approved in accordance with the applicable regulations for the

safe carriage of radioactive material adopted by the IAEA and carried in accordance with class 7

of the international maritime dangerous goods (IMDG) code, schedules 10, 11, 12 or 13.

For the purpose of the code, ships carrying materials covered by this code in flasks have been

assignedto three classes depending on the total radioactivity to be carried on board:

Class INF 1 - Ships carrying such materials with an aggregate radioactivity less than 4 000

TBq.

Class INF 2 - Ships carrying irradiated nuclear fuel or high level radioactive wastes with an

aggregate radioactivity less than 2 x 106TBq and ships carrying Plutonium

within aggregate radioactivity less than 2 x 105 TBq.

Class INF 3 -Ships carrying irradiated nuclear fuel or high level radioactive wastes and ships

carryingPlutonium with no restriction on the aggregate radioactivity of thematerials.

4.4.5-8

iL

All ships in addition, regardless of size, carrying materials covered by this code should comply

with Solas 1974 and the following requirements of the INF' code on:

Damage StabilityFire Protection

Temperature ControlStructural Considerations

Cargo SecuringArrangements

Electrical SuppliesRadiation Protection

Management, Trainingand Ship Board EmergencyPlan

BNFL is to classify its fleet of shipsunder the requirementsof INF 3.

The area of physical protection will require agreement between the appropriate UK and US

authorities on responsibility. Thereare a numberof contributingfactors such as type and country

of registration of vessel.

In the absence of any definiterouting requirementsit is not possible to determineany port specific

problems. However, "friendly"ports will need to be established.

4.4._9

4.4.5.4 Physical Protection Requirements

The UK has physicalprotectionplans/systemsthat:

a) complywithinternationalstandardsfor protectingnuclearmaterialin transit,and

b) are approvedby the relevantgovernmentagencies.

The internationalstandardsare:

INFCIRC/225/Rev 2: "The PhysicalProtection of Nuclear Material" and

INFCIRC/274/P,,ev1""Conventionon PhysicalProtection of Nuclear Materials"

to which the USA is also a signatory. Figure4.4.5-1 and 4.4.5-2 refer.

The principal requirementof a nuclear industryphysical protection system is to protect special

nuclear material against theft and radiological sabotage. Additionally, as a consequence of

obtaining an approved physical protection plan, that document would become a guide for the

implementationof the physical protectionregime.

The provisionof physicalprotection equipment and proceduresis dictatedby the level and nature

of the "threat" situations attributed to a process, plant or transportation system under

consideration. Detailed risk evaluation techniques exist within specialized security organizations

who, byvirtue of their knowledge, experience and relationshipwith Governmentdepartmentsand

agencies, are able to evaluate and quantifyphysical protection proposals against the "threat". An

acceptable physical protection plan has to comply with the regulatoryrequirementsoutlined in ---

above. A typical document hierarchyis shown in figure 4.4.5-3.

A. Communications

The communication network needs to be established which will define both routine and

emergency channels together with key individuals who would be involved throughout the whole

transport operation. Appropriate hardware on land vehicles.and ships will enable contact to be

maintained with an Operations Control Center.

B. Safeguards.and Security

This is an important part of the total documentation. Its purpose is to demonstrate that

satisfactory predetermined plans involving equipment and personnel (supported by back up forces

4.4.5-10

if required)can adequatelyrespondto safeguardemergencyevents.

A point to be resolvedherewill be the authorityandresponsibilityof the safeguardsand security

forces assigned to these shipments. At present, within the US, this authority and responsibility

restswith the USDOE and individualsselectedto providethe serviceare deputizedas Federallaw

"enforcementagents. Authorityand responsibility,forsea transports.wiUneed to be defined.......

C. Plannina

Thereare two aspects to the planningoperation. The first-involvesthe overallcampaignplanning.

The objective of this is to decrease vulnerability,and maximize the ability of any response forces.

To fulfill this objectivethe resultant informationfeeds into the safeguards contingency plan. The

second involves the planning for each movement which uses a codification system for sensitive

informationbeing passed between those parties involved.

Marine and land contingency plans provide essential information on sea and land transport,

engineering and health physics. This information is required to mount a response to a

conventional or radiologicalproblemon a ship or vehicle.

D. EmergencyResponseArrangements

Emergency response arrangementshave been developed for sea and land transport. Procedures

have been developedand refinedas a result of regularrealistic emergencyexercises.

Sea Transport

Although there is a high level of safety resultingfrom the PNTL ship design, it is necessary to

nlake contingencyplans for a major incidentinvolvinga nuclear fuel carrier. A full world-wide

emergencyresponse systemincludes:

a) round the clock expert adviceavailableto the ships masters;

b) ship position, headingand speed reported automaticallyto the operations center;

c) emergency response team on standbyat all times;

d) specializedequipment available;

e) equipment andtechniques developed from regular realistic emergency exercises;

f) world-wide salvagecover;

g) salvage location and telemetry system assists in vessel location and provides informationto the salvageteam regarding the condition of the ship and cargo.

4.4,5-11

LandTransport

Procedureshave been developed for emergencyresponse purposes for land transport. As for sea

transportthese have been refinedas a resultof regularemergency exercises.

Forland transport purposes the experiencegained fromthese exercises, together with the existing

emergency response documentation, would be used as the basis for defining and developing

appropriate arrangementsfor use in the US.

4.4.5-12

4.4.5.5 Licensing and Safeguards

Due to the lack of specific detailsit is not possible to answerthis directly,however the following

generalconditionsapply:

.... Regardingexport to the UK, itwould be,necessaryto know..whether,ornot the_material

would be undersafeguards alreadyin the USA and if not whetherthere was a need to keep

it out of safeguards on import to the UK. If there is no reasonto bringit into safeguards

then the US/UK Defense Agreement-couldbe used to effect the transfer (if it was .for

militarypurposes). If the material was to be used for non-nuclearpurposes then it could

possibly be brought into safeguardsand then exempted. If it is alreadyunder safeguards,

and its intendeduse is nuclear,then it will have to come underUS/EuratomAgreementfor

Co-operation,which will meanit will at all times be subjectto US controls.

As far as export licensingis concerned,a US export license will be needed irrespectiveof

safeguardsstatus. Only the authorizingagency will differ, dependingupon whether it's a

civil or non-civilpurpose. We understandthat licenses are issued at federalratherthanstate

level, therefore these should not change asa result of exports from different states. _-

We would also need to establish from the US whether they need a UK import certificate from

HMG before export is allowed as proof that the UK is in a position to receive the material. We

are unable to comment at this stage, on the conditions which would apply within or between

states in the USA as these may differ. Further advice needs to be sought to provide

comprehensive guidance on both State and Federal laws.

Overall, the biggest obstacle may not be the legal/regulatory framework within which exports take

place, but the political perspective within the US as to whether such movements are acceptable.

4.4.5-13

4.4.5.6 Routing Requirements

Without knowing specific destinations, definitive routing requirementsobviously cannot bedescribedhere.

....... In 'generalinstructionsto sailor-travel a particular.route,or _segment._would.be.transmittedby

t_ secure means. Information on the route and the _itinerary_, _ " _edas Confidential

Information in accordance with national guidelines.

Emergency arrangements,e.g., ports of call for ships,-vehicle breakdown/incidentsupport, etc.

would need to be established in a security contingency plan. These arrangementswould have to

be pre-determinedand appropriatepre-notificationrequirementsdefined.

4.4.5-14

INFCIRC/225/REVI EUROPE UNITED STATESNRC 10CFR

6.1.2(a) Shortest Journey Time / /6.1.2Co) Minimize Transfers / /6.1.2(c) Avoid Regular Schedules / /6.1.3 Code Names / /

6.2.1 Advance Notification / /6.2.2 Advance Authorization / /6.2.3 Selection of Route / /6.2.4 Locks and Seals / /6.2.5 Search of Vehicle / /6.2.6 Written Instructions / /6.2.7 Measures after Shipment / /6.2.8 Communication / /6.2.9 Emergency Action / /6.2.10 Escorts or Guards / /6.2.11/2 Advance International / /

Agreements

Mode Specific Provisions

6.3.2 Road / /6.3.3 Rail / /6.3.4 Sea / /

Figure 4.4.5-1. Physical Protection Requirements

4.4.5-15

IAEA UNITED KINGDOM UNITED STATESINFCIRC/225/REVI MINIMUM STANDARDS NRC 10CFR

Load to Secure YES ANSI or ISO

CompartmentContainer

One or more Guards: Case-by-Case YESArms not specified Provisions Armed

Type of Vessel: British Vessel Container ShipNot specified British Master

Minimi_ ports of call YES YES

Locks and Seals: YES To be InspectedInspected regularly Inspected regularly "whenever possible"

Communications: Case-by-Case Reports everyNot specified Provisions 6 hours

Figure 4.4.5-2. Physical Protection Requirements Special Provisions by Sea

4.4.5-16

Pu/MOX SHIPMENT DOCUMENTATION

hd TEROPEN TRANSPORTDCX_UMENT PLAN

II' IJ, ,- 1 J IMATRIX PLAN

• 3_dTEn . - .........(,_r_ort I

- I _L I,.,,,..lr DOCUMENTS)

PRE-SHIPMENT SYSTEMS TO PREVENTION OF TRANSPORT HEALTHPLANNING | COUNTER UNAUTHORIZED CONTINGENCY PHYSICS UNDER EMERGENCY HEALTH

DOCUMENTATION I ENTRyUNAUTHORIZEDDECEITREMOVALBY RESPONSEMANUAL MANUAL ESCORT ROUTE SITUATION PHYSlC,_MANUAL

1 ........)ETECTION OF ENGINEERING ENGINEERINGJNAUTHORIZED | EMERGENCY EMERGENCY_EMOVALBY / RESPONSE SITUATION RESPONsEEMERGENCY

;TEALTH OR FORCEI MANUAL AT LOCATION MANUAL/

TYPICAL DOCUMENT, SYSTEM HIERARCHY

5.0 SAFETY AND ENVIRONMENTAL APPROVAL

5.1 PU DISPOSITION COMPLEX S/_"ETYAPPROVALWITH TRITIUM PRODUCTION

5.1.1 Fuel Licensing Considerations

The ABWR certification provides for the operation ofalternate fuel designs provided that specific criteria onthe fuel and control blades are satisfied (NEDE 234011-P-A,Ammendment 22). These criteria evaluate not only the

thermal/hydraulic performance under transient and accidentconditions, but also stability considerations, impact on the

power/flow map and operating limits. NRC approval of eachreload is not required, provided that documentation that thecriteria are satisfied is available. NRC normally recievesa copy for information.

The ABWR SAR is based on the application of 8x8 UO 2 fuelrods. Licensing considerations for the initial MOX core isdiscussed in Section 2.3.1. Such considerations include

applicability of calculational methods and results of theanalysis.

Assuming that the MOX core has recieved prior NRC approval,operation for tritium production may still require anadditional specific submittal for NRC approval prior toinitiation of operation for the first production cycle.Additional study is required to determine whether theinitial tritium production core could be loaded withoutseparate NRC approval.

5.1.2 Potential Design Modifications

No design modifications to the ABWR have been identifiedspecifically required for the initiation of tritiumproduction or the disposal of plutonium. Therefore, theplanned certification of ABWR design satisfies the safetyapproval of the ABWR-PDR part of the complex.

5.1.3 Probabilistic Risk Assessment Impact

The Probabilistic Risk Assesment (PRA) contained in the ABWR

Safety analysis (Reference 3, Chapter 19) is based upon thecertified ABWR design. Since no design modifications arenecessary for Plutonium Disposal, this analysis remainssatisfied. The different core design and tritium productionlead to some other differences which are discussed below.Detailed review of differences should be conducted toconfirm that the PKA conclusions remain valid.

D__ecay Heat

Section 3.4.4.3 indicates that for both the plutonium

disposal cycle and tritium production cycles, decay heatwill be lower than for the core evaluated in the PRA.

Therefore, the current evaluation bounds the ABWR-PDR case

and no new analysis is necessary.

Fission Product Inventory

Section 4.5.1 of reference 1 provides a discussion of severeaccident source terms applicable to the PRA, but does notinclude a discussion of the fission product inventory. Table2.7-3 of reference 1 provides the inventory of spent fuel 5days after reactor shutdown, but does not provide anisotopic breakdown of fission products. The fission productinventory of the fuel assumed in the ABWR PKA is summarizedin Table 2A-6 of reference 3. Comparison of theseinventories with comparable values from the MOX core areneeded to confirm that bounding inventories of fission

products exist in the core evaluated in the PRA. Nosignificant differences are expected.

Fuel Pool Assist

RHR operation following startup in the Tritium Cycle willrely upon the Fuel Pool Cooling assist mode of the RHRsystem. In this operating mode, the RHR is not availablefor automatic initiation in low pressure core flooding modeor the suppression pool cooling mode. To satisfyoperability requirements of the Technical Specifications,however, manual realignment mode of the system may be

relied upon in the event that low pressure injection or poolcooling is required.

The potential for manual realignment has a slight impact onthe core damage frequency evaluated by the ABWR PRA.Sensativity studies indicate that core damage frequencywould be increased by much less than 10%. Therefore, thisimpact, although it represents an increase in risk, can beconsidered insignificant.

5.1.4 Licensing Considerations for the Disposition

Complex

Safety regulators, and therefore assumed organizationsresponsible for requirements for the disposition complex,are summarized in the table 5.1-1. Licensing considerationsfor DOE-Defense and waste management are not considered in

this phase of the project since such activities are assumedto be beyond the boundaries of the facilities underconsideration. Only the Fuel Fabrication and PowerProduction portions of the complex are addressed in thisreport.

Application of Certified ABWR Design to the ABWR-PDR

Assuming the NRC licenses the operation of the powerproduction complex, any ABWR design changes required by thePu/Li production must be evaluated against the Inspections,Test and Acceptance Criteria (ITAACs) to establish the basisfor safety approval of the ABWR-PDR portion of the complex.Since no modifications, other than the fuel design, havebeen identified, the certified design is considered to beintact for the ABWR-PDR.

In addition to submittals for the fuel design, certain sitespecific data requires approval to complete requirements foran operating license. Further, a discussion of NEPA actionsrelative to the ABWR-PDR plant as well as the fuelfabrication facility, must be considered for final NRCapproval. Additional discusion is provided in Section 5.2.

The need for initiation of a tritium production cycle withinsix months of notification places scheduling considerationson the licensing of the tritium production core. Priordesign and approval of this core would be prudent so thatonly hardware and operational considerations will need to betaken into account.

DOE Requirements

Section 4.2 of the Reference 2 summarizes the applicablerequirements for the DOE regulated portion of the complex.A regulatory line of responsibility will need to beestablished between NRC and DOE for approval of the complex.A concept to approach approval is described in Section 5.2,but the ABWR certification effort and basis needs to be

integrated into the concept.

No specific regulatory requirements have been identifiedwhich apply to the target rods or handling of exposedtargets other than DOE and NRC requirements associated withthe handling of radioactive materials. In order assure thattargets remain intact during operation and handling,fabrication quality controls are expected as a purchaserequirement, but not subject to regulation. Section 3.3provides additional information on target fabricationrequirements.

5.1-3

5.1.5 References

1. "Study of Pu Consumption in Advanced light WaterReactors", GE Nuclear Energy, NEDO-32292, May 13, 1993.

2. "Study of Pu Consumption in Advanced light WaterReactors, Compilation of Phase ib Reports", GE NuclearEnergy, NEDO-32293, September 15, 1993.

3. "Safety Analysis Report, Advanced Boiling WaterReactor", 23A6100 Rev I, Amendment 31.

Table 5.1-1

Plutonium Disposal ComplexRegulatory Agencies

II llrl

Disposition Complex Segment RegulatorWeapons Reciept Facility DOE-Defense

Target Fabrication Facility DOE-Defense

MOX Fabrication Facility DOE-Nuclear EnergyPower Production Facility NRC(ABWR-PDR)

Tritium Recovery Facility DOE-DefenseWaste Storage Facility DOE-Waste Management

, ..... , ,,

5.1-5

5.2 IMPACTOF TRITIUIt PRODUCTIONON ENVIRONMENTALAPPROVAL

As discussedin the Phase IA report (ReferenceI) the only incrementaleffects

of tritiumproductionon environmentalapprovalare relatedto the leakageof a

small amount tritium to the environmentfrom the ABWR. Since only one ABWR is

necessaryto achieve the tritiumproductiongoal, thls assessmentassumesonly

one tritium-producingreactor.

Currently,there are no federalor state regulationscontrollingthe releaseof

hydrogen-3 (tritium) to the environment with respect to chemical hazards

(hydrogengas or water/watervapor). However,becausetritium is a radioactive

isotopeof hydrogenthat could cause radiationdose to the public,releases to

the environmentare regulatedunder severalfederaland state statutes.

State regulationspertinentto this issuevary from state to state,but generally

follow the federal regulationsvery closely. For that reason, only federal

regulations will be considered in this discussion. The applicability of

regulations also depends upon the specific approval authority to operate the

facility. A facility operatingunder DOE authoritymust comply with standards

issuedby both DOE (DOEOrders) and the EPA (Code of FederalRegulations,Title

40, EnvironmentalProtection). A facility licensedby the NRC must be capable

of demonstratingcompliancewith standardspresentedin Titles 10 (Energy)and

40 of the Code of FederalRegulations. A brief descriptionof the dose limits

is as follows"

Departmentof Enerqy

DOEOrder 5400.5

• 100 mrem/year effectivedose equivalent (ede) from all DOE sources and

exposure modes

• 10 mrem/yearede from airbornecontribution(by referenceto 40 CFR)

4 mrem/year ede from drinkingwater pathway

5.2-I

• Best Available Technology for ltquid discharges to surface waters

• Best AvailableTechnologyfor liquid dischargesto sanitary sewers

40 CFR61 subpart H (National Emission Standards for Emissions of Radtonucltdes

Other than Radon from Department of: Energy Facilities)

• 10 mrem\yearede from airbornecontributionfrom entire site

NuclearRequlator.YC.ommission

10 CFR20

• 100 mrem/year total effectivedose equivalent (excludingsanitary sewer

releases)

• Annual average concentrations do not exceed Table II (10 CFR 20

Appendix B) limits at the unrestrictedboundary

• Annual external dose not to exceed 50 mrem at the unrestrictedboundary,

continuously occupied area

40 CFR 61 subpartI (NationalEmissionStandardsfor RadionuclideEmissionsfrom

Facilities Licensed by the Nuclear Regulatory Commission and Federal

FacilitiesNot Coveredby SubpartH)

• 10 mrem\year ede from airborne contribution (all radionuclidesexcept

radon) from entire site

• 3 mrem\yearede from airbornecontribution(radioiodines)

Basically, there are no regulatory requirements specific to tritium. For

environmentalapproval purposes, the impact of tritium on the environmentis

included as a component of the total radionuclideimpact. For this report

5.2-2

however, the impact of tritium released to the environment as a result of tritium

production activities ts assessed as an incremental impact.

This report ts a refinement of the Phase 1A assessment on tritium impact to the

environment. That assessment determined that, when the Incremental tritium

Impact was added to the other radtonucltde impacts, the.total impact was within

the standard levels specified above. If the impact of tritium is considered

alone, the impact was projected to be a_factor of 25belowtheregulatorylevels.

In actuality, the impact would be much less without the very conservative

assumptions used in the first assessment. Thts section introduces and discusses

those refinements that bring the impacts closer to reality.

[nvironmentalImpacts

The refinementof the tritiumconcentrationinthe reactorcoolantin Section3.2

effects the quantity of tritium released to the environmenteach year during

normal operations. Table 5.2-I provides a listing of the revised tritium

environmentalreleasequantities. These quantitieswere used to determinethe

environmentalimpact of tritiumproductionusing the ABWR.

Table 5.2-1. Tritium Release Quantities, ;, , , ..... , . ,, ,,,,, _ --., ,,

Tritium Source Operational Period Annual Quantity

Released (Ct)'1m i ' iiiii i iiiii i __ i imli i

ReactorBuilding Dur.!ngPower 30.6

During Refuel!ng 24.3

Turbine Building Dur!ng Power 346.9

Total 401.8

By far, the most restrictiveregulation pertaining to the tritium production

activity is the 10 mrem/year ede limit to the maximally exposed person.

Therefore,the impactof this activitywill be assessedagainst the regulatory

limit of 10 mrem/year. This limit is found in 40 CFR 61 in both subpartH (DOE

Facilities) and in subpart I (Nuclear Regulatory Commission -licensed

5.2-3

facilities). These regulations are better known as the National Emlsston

Standards for Hazardous Air Pollutants (NESHAPs).

These impacts were assessed using two calculattonal methods approved by the EPA

for documenting compliance to NESHAPSstandards. These are AIRDOS-PC(Version

3.0, November, 1989) (Reference 2) and CAP88-PC (Version 1.00, 1989)

(Reference 3). Due to the absence of site meteorological data, worst case

meteorology was assumed. Where the two codes gavediffering results, the highest

calculated value was used.

No liquid dischargesare expectedto containtritium;only the gaseouseffluent

from the plant is expectedto containtritium. Chemically,the releasedtritium

is assumed to be within water molecules. This assumption gives the most

restrictivevalues for healthdetrimentfrom exposure. Ingestionand inhalation

are the only two significantpathwaysintothe body and the only ones considered

in this assessment. These pathways result in only an internalradiationdose.

AIRDOS-PC determined the maximallyexposed member of the public would receive

0.0092 mrem/year ede from tritium at ]000 meters downwind from the facility

stack. This is at least a factor of 1,000 less than the 10 mrem/year limit.

However, because the airborne dose contributionfrom the entire site must be

included in the 10, no conclusionscan be drawn except that tritiumproduction

should have a very minimal impacton the environment.

When put into a risk perspective,a radiation dose of 0.0092 mrem ede would

relate to a lifetime fatal cancer risk of 3.7 E-9 for the maximally exposed

individual. Ariskto the entireexposedpopulationcannot be projectedwithout

offsite populationdata.

The Phase IA report assessedthe tritium impact at two potentialsite boundary

distances. These were at 400 meters for a standardNRC-licensedcommercialpower

plant and at 10,000 meters for a typicalDOE Site boundary. The CAP88-PC code

determinedan annualdose of 0.0014mrem ede at the 400 meter distanceand 0.0028

mrem ede at the 10,000 meter distance.

5.2-4

OccuDatlonal Impacts

Therefinement of the tritium concentration in the reactor coolant in Section 3.2

also effects the quantity of tritium released tnto the ABWRfacilities eachyear

during normal operations. This in turn dtrectly effects the occupational

exposure to factltty workers. Table 5.2-2,provtdes a ltstlng of the,revisedfacility tritium airborne concentrations. These quantities were used to

determine the occupational impact of-tritium production using the ABWR......

Table 5.2-2. Facility Tritium Air Concentrations

Plant Location Operational Period Air Concentration:c

Reactor Building Power 7.84 E-08

Duri Refuel i n_ 1.86 E-07Turbine Build' ,er 1.55 E-07

As can be seen, the concentrations are significantly below the derived air

concentration (DAC) of 2 E-05 pCi/cc and are an order of magnitude below the

tritium concentration (2 E-06 pCi/cc) that DOE requires for respiratory

protection of workers. Theseconcentrations are an average over the room air

volume. It is possible that, near the sourceof the tritium entry into the air,

concentrations could be higher. However, it is unlikely that workers would be

exposedto concentrations that would require respiratory protection.

Facilityworkerswouldbe exposedto the Table5.2-2air concentrationsfor a

limitedperiodof timeand ina limitednumberof locationsin theplant. Within

the reactorbuilding,workersshouldbe exposedonlyduringrefuelingactivities

where tritiumatoms becomeairbornein water moleculesfrom evaporationof

coolantwater. In the turbinebuilding,tritiumbecomesairborneduringpower

operationfromminorsteamleaks. Theseare the only two areaswheretritium

contaminantsare expected.

5.2-5

From the projected concentration levels and the anticipated annual occupancy in

each area, the worker radiation dose from tritium can be determined. Table 5.2-3

lists the collective dose to workers from tritium airborne contaminants. The sum

of these doses total 0.184 rein, which relates to a risk of 7.4 E-5 for incurringa latent fatal cancer.

To placethis dose in perspective, it tsassumedthat four.workers receive the

entire dose from tritium exposure (4 workers at 2000 hours per year ~ 8,500

person hours). The average dose to an individual worker is 0.046 rem per year.

This maximumdose would be two orders of magnitude less than the DOEannual limit

of 5 rems and is a factor of 12 less than the DOEannual dose goal of 0.5 rem.

This worst case scenario indicates no significant impact to occupational workers

from tritium production.

Table 5.2-3. Incremental Occupational Doses from Tritium Productioni

Location Person Tritium DAC EDEHours Concentration Fraction (Rem)

ci/cc , iill ii i

Reactor Bldg. - 5,000 1.86 E-07 0.0093 0.116Refuelin_

Turbine Bldg. - 3,500 1.55 E-07 0.0078 0.068During Power,,, i

Total 8,500 0.184

Summary

This sectiondeterminedthat the maximallyexposedoffsiteperson could receive

a radiationdose of 0.0092mrem per year fromABWR tritiumproductionactivities.

This is a very small percentageof the EPA dose limit of 10 mrem ede per year.

The impactto occupationalworkers is projectedto be a collectivedose of 0.184

rem per year. Spread over several facility workers, this is a very small

percentage of any occupationaldose standard. Therefore, with respect to

environmental,health,and safety impact,tritiumproduction in the ABWR would

presentno technicalbarrierto projectapproval.

5.2-6

References

I. Study of Pu Consumptionin Advanced Light Water Reactors- Evaluationof

GE Advanced Boiling Water Reactor Plants,NEDO-32292 May13, 1993

2. U.S.E.P.A, User's Guide for AIRDOS-PC, Version 3.0, EPA/520/6-89,035,

December, 1989, Office of Radiation Programs, Las Vegas, NV.

3. U.S.E.P.A., User's Guide for CAP88-PC, Version 1.0, 402-B-92-001, March,

1992, OfFice of Radiation Programs, Las Vegas, NV.

5.2-7

5.3 ABWR PU DISPOSITION COMPLEX SAFETY APPROVAL PROGRAM

5.3-1 Program Assumptions

The safety approval program assumed for the ABWR Plutonium Disposition Program willconsist of the submittal of a General Electric Integrated Safety Analysis Report (ISAR) and aDOE Integrated Safety Evaluation Report (ISER) safety review process. This programassumes that DOE wiI1 b¢ responsible for the overall safety approval program with input asrequired from other government safety agencies, committees or boards such as the NuclearRegulatory Commission (NRC), the Defense Nuclear Facilities Safety Board (DNFSB), or anadvisor5' board on Pu Disposition safety similar to that established for the New ProductionReactor (NPR) program.

It is also assumed that the ISAR/ISER documents will cover the entire ABWR Pu Dispositioncomplex including the mixed oxide (MOX) fuel fabrication facility and DOE approvals willallow early start of construction and operation of the mixed oxide (MOX) fuel fabricationfacility before construction/operation of the ABWR reactor facilities.

5.3-2 Integrated Safety Analysis Report Development and Submittal

The Integrated Safety Analysis P.,epon will serve as the single, primary safety evaluationdocument provided by General Electric for the Pu Disposition Complex. This document willcontain all design and analysis information required for review of the safety adequacy ofthe ABWR reactor, MOX fuel fabrication tad complex support facilities.

The ISAB, will be developed and submitted in a phased process to support DOE comprehensivereviews and conclusions that are necessary for readiness reviews conducted by DOE at keydecision points established for authorization to proceed to the next program phase. Asshown in Figure 5.3-1 these decision points include authorization to proceed with Title IIdetailed design, to begin site preparation activities, and to proceed with construction,startup and operation. Each ISAR submittal required to support these key decisions isbriefly described below.

• ISAR Submittal 1- Identification of Safety Requirements and Criteria andEvaluationMethodology- consists of ISAR sections describing the safetyrequirements and criteria and evaluation methodologies that will bc used to verifythat the required level of safety has been achieved.

• ISAR Submittal 2- Initiation of Site Preparation Activities- this submittal willsupport the start of site preparation activities and will include preliminary designinformation for all safety systems, seismic, meteorologic, hydrologic tnd geologiccharacteristics of the site, anticipated maximum levels of radiological and thermaleffluents the complex will produce, proposed major features of the emergencyresponse plan, and the level I probabilistic risk assessment,

• ISAR Submittal 3- Authorization for Substantial Construction- will integrate all thepreviously submitted material into the complete ISAR tad include information that issufficiently detailed to permit DOE to reach definitive safety conclusions. In addition,the Construction Safety Verification Plan (CSVP) including the principal verificationinspections, tests and analyses and acceptance criteria will be provided, The ISAR, atthis stage, will necessarily have less detail than contained in commercial nuclearpower plant operating license applications, but wiI1 be more detailed and completethan a typical application for a construction permit, The basic level of dotal1 isshown in Figure 5.3-1.

• ISAR Pro-Operational Amendments for Authorization to Load Fuel and PerformStartup Testing- these preoperational amendments will provide detailed informationnecessary to demonstrate the readiness for plant startup testing and the conduct of

5.3-1

operations. Revisions to previously submitted design and analysis material willreflect the as-built complex and will include the final complex technicalspecifications as well as the final accident management and emergency responseplans.

• ISAR Amendments to Support Complex Operations- amendments will be provided tosupport resolution of outstanding issues required for the transition to productionoperations. The information submitted is expected to be minimal at this point, assafety issues will have been resolved early in the program, the as-built plantreconciled against the approved design, and operational issues addressed prior to fuclload and startup testing.

5.3-3 DOE Reviews O,,'ersight Reviews and Integrated Safety Evaluation Repon

Based on the submitted ISAR material, review and audit of additional engineering and safetyanalysis material, and answers to initial DOE questions, the DOE will issue drafts ofIntegrated Safety Evaluation Report (ISER) segments in the sequence of ISAR issuediscussed above. These draft ISERs will describe the DOE safety review and conclusions,including any conditions and confirmatory items necessary to verify safety features• Thedraft ISER segments will contain open items that require resolution prior to issuance of theISER. Input to the ISER will include results of oversight reviews by safety oversightagencies _nd committees such as the NRC and the DNFSB.

5.3-4 Safety Verification Program

A Safety Verification Program (SVP) will be established to identify, document and trackopen and confirmatory items that result from the generation and review of the design andsafety analysis information contained in the ISAR and to track to completion the tests,inspections, analyses and acceptance criteria used to verify the as-built plant confomlswith the approved de.sign. This program will include a Design Safety Verification Program(DSVP) and a Construction Safety Verification Program (CSVP). The DSVP will beimplemented during the design and analysis phase of the project and will include trackingand closeout of open items requiring resolution as well as genetic industry issues arising inthe commercial nuclear power industry or from other DOE nuclear production facilities.The CSVP will be used to document and track plant construction and performanceconfirmatory tests, inspections, analyses and acceptance criteria used to verify the the as-builtplantis in conformancewith the approveddesign.

5.3-5 ABWR ReactorSafetyApprovalProgramSchedule

The ABWR reactorsafetyapprovalprogramscheduleisshown on Figure5.3-2The ISARsubmittalsare phasedto supportthe DOE safetyreviewprocessforISER developmentand tosupportkey program decisions.Durationsshown presentan aggressiveschedulebased onthenearlycompletedreviewof the GE AI3WR StandardSafetyAnalysisReport(SSAR) by theNRC. The ABWR isscheduledtoreceivefinaldesignapprovalfromtheNRC in Ib'94.Also,theABWR First-Of-AKind Engineering(FOAKE) forthe standardU.S.plantdesignhas beenstar_ed In addition, two ABWRs are currently under construction in Japan and are scheduledfor operation in 1996 and 19_1, This considerable experience in design, licensing andconstruction of the ABWR gives a measure of confidence in the proposed schedules.

5.3-6 ABWR Pu Disposition Program MOX Fuel Fabrication Facility Safety Approval Schedule

As shown in the overall ABWR Pu Disposition Program Schedule (Figure 5.3-3 ) theauthorization for start of construction of the MOX fuel fabrication facility is scheduled atapproximately the 34th month after start of design versus the end of the 36th month for theABWR reactor complex. This schedule is predicated on expeditious development of facilitydesign and early environmental and safety approvals. Again, the schedule for this facility is

5.3-2

considered aggressive but can be achieved with a national commitmcnf and proper use ofexisting technology both in the U.S and abroad. As noted ia Sections 2.2 and 4 thetechnology and infrastructure for MOX fuel fabrication at both DOE and foreign sitescurrently exists and can be drawn upon to expedite the schedules. Although licensingapprovals of some modern facilities in other countries have been extensive, it is believed thatthere are no unresolved technical issues that would delay approval of the safe operation ofthis facility on an environmentally acceptable DOE site,

5.3-3 iII

STARTTITLEII BEGINSITEPREP AUTH.C_RUCTION STARTUP OPERATION

ISARSUBMITTAL1 ISARSUBMITTAL2 ISARSUBMITTAL3 PRE-OPAMENDM'I"S AMENDMENTSi i l =

|ill

' • Safety requirements • Tille I safety • Generalarrangement - Conductofop's - Resolveopen itemsi andcriteria systemdesigns

• System functional• Analysismethods - Site interface descriptions * Summaryof startup • CompleteCSVPtests

• P&lDs• Radlologicaland - Final E-Plan

thermal effluents • Controllogicdiag.'s

u_ • Major featuresof • ALARA/radlation -Tech specs;,,,,J= E-Plan protectionprogram - CSVP.resulls

• Security plan• Lever1 PRA• Accidentmgmt plan

• Prelim. E-Plan

• DBA's

• Severeaccidentassessment

• Prel. level 3 PRA

• Drall tech.specs

• CSVP inspections,lests,analyses

,, ,,

FIGURE 5.3-1 ISAR SUBMITTAL SEQUENCE & GENERAL CONTENT

Months 6 12 18 24 30 36 42 48 54 60 (;G 72 78 84 90 93

START I ROD SEE EIrARI"PREP CONSTRUCTION FUELLOAD OPERATION

ENVIRONMENTAL EiS ] I ,_L,co_s'reuCnON I I

SAFETY I i Y AUT,ORIZA_ON I I

ISAIVISER I I I I I

SAFETYCRITERIA ISARsue ! ISERI I I ! I

sITEPREP. i tSARSL_Z ,, / wseez I I I I

CONSTRUCTION ilSAR SUlB3 ._ ISER3 I I i

AMENDMENTS [ AMENDMEN_$ ' "' • / PRE-OP' 1 1

SAFETYVIERIF. I I | I IL_

I I _ I IDESIGN(DVSP) [ DESIGN/ANALYSISOPENITEMS l

I I I I I

CONSTRUCTION(csvP) [SUBMITCSVP i JREV_WCSVP I PERFORMCON:>rlRUCTJONiPERFORMANCEVERIFICA31ON , [

t i _ I IENGINEERING lilLE I hTLE !! I COMPLETETITLE !1 J

i I I I I

PROCUREMENT l, . . PROCUREMENT I I I

CONSTRUCTION ] , i IlilLE nlSUPERVISION i I I i

I SITE_I I ICONSTRUCTION!, PREP I CONSTRUCTION t ]

TESTING/ I ISTARTUP I _E-OPTESTS SrA_TUP1

FIGURE 5.3-2 ABWR REACTOR PU DISPOSITION SAFETY PROGRAM SCHEDULE

' Activities / Year

DEVELOPMENT

SAFE'rY/ENVIRONM.

EIS

-- ConslructionSAFETYVERIF.

FUELI rCENCING

REACTOR - Title 1DEIGN

Site ConstructionOONSTRUCnON

PRE.OPTEST

OP.READINESSREV_

_, STAFTI1JPL_

OPF.RAllON

FUEL FAB PLANT

DESIC__

_ON

PRE-OPTEST

OP.REAOINESSREV.

STARTUP

OPERATION

MILESTONES

Rx Env. SI art- Reac- ReactorTil. I RE_ of torCorn- Fuel Fuelplele F'ab load

,, IIFIGURE 5.3-3 ABWR PU POSITION PROJECT SCHEDULE-ONE

5.4 ENVIRONMENTAL PERMITI'ING PLAN AND SCHEDULE

5.4.1 Introduction

This section contains a preliminary plan and schedule for obtaining theenvironmental permits and approvals required for the constructionand operation of the Advanced Boiling Water Reactor (ABWR) Pu-disposition facility at a "greenfield" site and at an existing DOE facilitysuch as the Savannah River Site (SRS). The Preliminary Schedulepresented here assumes that ABWR safety approvalsand permittingwill use Department of Energy (DOE) Orders and Regulations asprimary guidance but will also be compatible to those of the NuclearRegulatory Commission (NRC).

The Plutonium Disposition Complex (PDC) at either a greenfield site orat an existing DOE complex will consist of two co-located facilities: aMixed Oxide Fuel Fabrication (MOX) Facility and the Advanced BoilingWater Reactor (ABWR) and Power Block. The two generic sitingalternatives discussed will generally bound the schedule uncertaintiesfor the combined facility configuration.

A specific site should fall within these bounds with. the "degree ofdifficulty" of the selected site largely determined by site ownership andcontrol. In the case of the ABWR using Plutonium as fuel,implementation possibilities range from nearly impossible for a green-field site to quite possible at an existing DOE site such as the SRS,ORNL, or INEL.

At the preliminary schedule level of analysis used here, an existingDOE site offers significant schedule acceleration (two to five years),more certainty, and a lower level of complexity for the environmentalapprovals process.

The permitting process for the ABWR at an existing DOE facility wouldmostly bypass the site selection, evaluation, field data collection effort,and public involvement process required by a greenfield site. No fielddata would have to be collected. Non-nuclear permits would also beless complex and time consuming. In addition to providing a site withthe required security, skilled work force, and nuclear supportinfrastructure; a DOE site would have many required permits alreadyin place which could be modified to accept the ABWR.

Because of the extreme sensitivity of the public to siting nuclearfacilities, any greenfield site would require extensive publicinformation and relations programs to begin to have a chance to gainapproval. Some communities near existing DOE site are familiar with

5.4-1

DOE operations and are believed to be supportive of expanded siteoperations.

5.4.2 Assumptions

It is assumed that the general criteria for permitting the ABWR assumethat the facility should be licensable by the NRC even though it wouldbe DOE project and DOE rules and procedures apply...DOE and otherapplicable regulatory agency standards and codes that are morerestrictive than those of the NRC (10 CFR) are considered as applicable ....

Additionally, it is recognized that the environmental evaluationprocess for this facility will follow the requirements and guidelines ofthe National Environmental Policy Act (NEPA). The procedures forobtaining a project Record of Decision (ROD) and approval to proceedthrough the NEPA process, although not strictly a permit, requiressignificant time and resources to complete. The permits and approvalsnot strictly part of the NEPA process require a favorable ROD to becomeeffective and the conditions attached to these permits and approvalsare incorporated as enforceable conditions under the ROD.

The assumptions used to build the environmental permitting schedulepresented in this report are described below divided into four areas:general facility siting, environmental assessment process,environmental permits and approvals, and nuclear licensing andapprovals.

5.4.2.1 Facility Siting

1. It is assemed that the "Facility" will consist of the ABWR andpower block co-located with a MOX fuel fabrication plant.

2. In accordance with the DOE Pu Disposition RequirementsDocument the ABWR will be located at a "greenfield" site nearKenosha, Wisconsin.. For contrast, the ABWR could be locatedat an existing DOE facility which would provide a trained andexperienced work force, a nuclear complex supportinfrastructure and nuclear materials security from resourcesalready on site.

3. Because of the unique nature of the nuclear materials (weaponsgrade Plutonium) employed as fuel as compared to a commercialpower reactor, nuclear materials security is a primary concern.

5.4 -2

4. Generated electric power (-1300 MW) will be accepted by thelocal grid in either the greenfield site or the DOE site. The abilityof the local area to adsorb the ~1300 MW will be a siting criteria.

5. The permitting schedule will address two plant site possibilities:1) a "greenfield" site and 2) a specific DOE controlled site. Thegreenfield site will require a generic approach to permitscheduling as site specific factors effecting permitting are notdefined. The DOE site location will incorporate as much sitespecific information as necessary to define a preliminary permitschedule.

5.4.2.2 Environmental Impact Assessment Process

1. DOE will be required to prepare two Environmental ImpactStatements: 1) a Programmatic EIS for the ABWR/PuDisposition Concept and 2) a site specific EIS for theABWR/MOX fuel fabrication facility at an existing DOE facilityor at a greenfield location.

2. The preparation of these two EIS's can proceed in parallel, not inseries. Other site specific EIS's could also be prepared as required.

3. The EIS process (one or more) ;vill take a maximum of two yearsto complete (for an existing DOE site) and four to five years for agreenfield site.

4. Environmental background data for candidate DOE sites hasalready been collected and analyzed and will be made availableto the ABWR EIS team as required. Extensive field surveys willnot be required. In contrast, a greenfield site will requireextensive data collection efforts to support an EIS(s).

5.4.2.3 Environmental Permits and Approvals

1. It is assumed that the environmental permitting process willaddress the facility as a whole as described in the EIS(s).

2. Sufficient design information exists and engineering help isassumed to be available to support the various permit andapproval applications either at a greenfield site or at an existingDOE facility. A green- field site will require the collection of sitespecific data.

3. DOE will assume the role of Lead Agency and will be responsiblefor inter-agency coordination.

5.4-3

4. The Preliminary Permit and Approval Schedule will addresscurrent regulations only. Impending and future regulatoryactivity will be incorporated with time, as the PreliminarySchedule is reviewed and revised to incorporate DOE experienceand judgement.

5. The Preliminary Permit Schedule will incorporate estimates forpublic review and comment of the EIS(s) determined fromexperience. The schedule will assume a beneficial effect ontiming that will result from an effective Public Relations andInformation Program.

5.4.2.4 Nuclear Safety Approvals

1. DOE Orders and Regulations will be the primary requirementsfor licensing. NRC procedures will be applied as appropriate andthe NRC will be kept informed of all safety approval andenvironmental activities. Materials developed specifically forNRC review will benefit the overall and site specific NEPAcompliance process and expedite permitting.

2. NRC comments and concerns will be addressed by the ABWRenvironmental and EIS teams. Applicable NRC Regulatory andSafety Guides will be referenced and used as guidelines asnecessary.

5.4.3 Applicable Regulatory Standards

, Based on the permitting criteria and approach described above, thefollowing are the major regulatory policy/organizations/agencies andtheir standards that apply to the ABWR:

• NEPA - Section 102 (2) (C)

• DOE - Orders No. 5400.5, 5480.11, and 5820.2A

* NRC - 10 CFR 20, 10 CFR 51, 29 CFR

• EPA - 40 CFR 60, 40 CFR 61, 40 CFR 403 - 471

• STATE* - All applicable state regulations

• LOCAL*- All local agency standards

* - where the facility is sited.

5.4-4

This list is not complete or exhaustive, but sites the majorgovernmental agencies involved in order to indicate the depth andextent of requirements and coordination efforts necessary forcompliance.

5.4.4 Environmental and Permitting Requirements

5.4.4.1 Environmental Impact Statements (EIS)

The National Environmental Policy Act (NEPA) requires identification ....and assessment of impacts to the environment from all major projectsproposed, such as this plutonium disposition project. In order to fulfillthis requirement, the project will eventually require a comprehensiveEnvironmental Impact Statement (EIS), which will involve asignificant amount of information to be collected or developed forassessing such impact.

For a project of this type and magnitude, the following resources andissues must be addressed and characterized, for determining theproject's impact"

* Land / Geologic Resources

• Air Resources and Noise

* Water Resources

* Land Use, Recreational, and Visual Environment

• Biotic Resources and Endangered Species

• Cultural Resources

• Radiological Impacts

• Nonradioactive Hazardous Materials

• Socioeconomics

• Transportation

• Waste Management

• Decontamination and Decommissioning

5.4-5

* Decontamination and Decommissioning

NEPA requires a 90-day public comment period after publication of theDraft EIS in the Federal Register. During this period, no decision on theproposed action can be made or recorded. NEPA has a similarrequirement of 30-days after publication of the final EIS, prior to theRecord of Decision (ROD).

Realistically, the total time period, from start or Notice of Intent (NOI)to prepare EIS to obtaining a ROD, is two years on an average to fouryears or more for a controversial project.

5.4.4.2 Permits

The following are the individual permits that are needed for anABWR facility at an unspecified greenfield site. The permit list

represents a conservative estimation of the overall permittingrequirements for the facility. A DOE site, such as the SRS, would alsorequire a similar list of permits and approvals but many of these wouldbe modifications of existing site permits and easier and less timeconsuming to acquire.

• Section 404 ( Clean Water Act)/Section 10 (Rivers and HarborsAct) Permit

• National Pollutant Discharge Elimination System (NPDES)Permit

• Air Quality Construction Permit

• Section 401 Water Quality Certification

• Stormwater Permit

• Hazardous Air Pollutants Emission (NESHAP) Permit

• Permit for Construction/Operation of Domestic Wells

• Permit for Construction/Operation of a Public Water SupplySystem

• Underground Storage Tank (UST) Construction Permit

• Permit to Construct Sewage Treatment Plant

• Sanitary Landfill Permit

5.4-6

• Radioactive Waste Transport Permit

• Hazardous Waste (RCRA Parts A&B) Permit

• Notice of Construction/Alteration to FAA

• Radioactive Material Transport (DOT) Permit

• Permit to Construct Solid Waste Management System

A brief description of each permit follows, and a permit scheduledepicting the milestones in their acquisitions is attached (Figure 2).

5.4.4.2.1 Section 404 / Section 10 Permit

Generally, the U.S. Army Corps of Engineers (COE) is involved inprojects when construction occurs in a waterway or wetland. Anyproject which has the potential to discharge dredged or fill materialinto the waters of the U.S. is required to obtain a permit from the COEas authorized by Section 404 of the Federal Clean Water Act (CWA).Section 404(h) of the CWA allows transfer of administration of thispermit program to qualified states. There is provision for public noticeand opportunities for public input in the permitting process, as theregulatory purpose of Section 404 is to balance public and privatebenefits and interests against resulting impact on aquatic environment.

Section 10 of the Rivers and Harbors Act authorizes COE to regulateactivities and issue permits for projects involving construction innavigable waters of the U.S. after notice and opportunity of publicinputs, similar to Section 404.

Time requirement for review and approval is approximately one year.

5.4.4.2.2 National Pollutant Discharge Elimination System(NPDES) Permit

The NPDES permit is administered _ander Section 402 of the FederalClean Water Act (CWA) provisions promulgated by the U.S.Environmental Protection Agency (EPA). Under Section 402(b) of theAct, states can administer their own permit programs under thedelegated authority from the EPA, providing a regulatory framework toenforce standards for protecting water quality.

5.4-7

The NPDES Permit regulates the point source discharge of pollutantsinto the waters of the U.S. Typically, industrial discharges regulatedunder the NPDES program include process wastewaters, contaminatedarea drainage, and stormwater during construction. Criteria andstandards for the NPDES permit system are described in 40 CFR 125.The permit application requires information on water use, wastewaterflow, characteristics and disposal methods, planned treatment andimprovements, stormwater treatment, plant operation, material andchemical used, and other pertinent information. Depending on projectcomplexity, the processing time for an NPDES permit varies from 6 -12_months. A public hearing may be required.

The agency specifies conditions in the NPDES permit on issuance,which include technology-based effluent limitations for the waste-streams as well as water-quality based limitations for the receivingwater. New industrial facilities are generally required to meet bestconventional (control) technology (BCT) for conventional pollutantparameters (e.g., COD, BOD, TSS, pH, oil and grease), and best availabletechnology (BAT) for toxics and non-conventional pollutants. Thepermit conditions will include monitoring requirements andprovision for additional technical requirements to checkconformance.

5.4.4.2.3 Air Quality Construction Permit

For the protection of air quality, the Environmental Protection Agency(EPA) sets air pollution standards that apply nationally through CleanAir Act (CAA) and its subsequent amendments. Additionally, state andlocal governments , through air pollution control districts (agency),have broad responsibilities for implementing air pollution controlstandards and regulations within their jurisdictional boundaries.

Each proposed new or modified air contaminant source must undergoa new source review. As part of this review, PSD (Prevention ofSignificant Deterioration) applicability is determined. If PSD review isrequired (generally applicable for facilities emitting more than 100 tonsper year of a regulated pollutant, or as designated by the local agency), aPSD application must be submitted and a permit obtained beforebeginning project construction.

The air permit requires identification of all stationary sources in thefacility, type and amounts of pollutants produced, and air pollutioncontrol equipment used. The permit processing time (for review andapproval) ranges from 6 weeks (no PSD) to 6 months (with PSD)assuming no additional data collection. The construction permit asissued is generally in effect until the completion of construction, after

5.4-8

which the agency maintains compliance through the issuance ofoperating permits.

5.4.4.2.4 Section 401 Water Quality Certification

A Water Quality Certification is required for a Federal License orPermit to conduct any activity that may result in a discharge intosurface waters, pursuant to Section 401 of the Clean Water Act (CWA).The federal agency is provided a certification from the state that thesaid discharge complies with the discharge requirements of federal lawand the aquatic protection requirements of state law. Generally, it takes4 - 6 months to obtain the Water Quality Certification.

Activities requiring this certification include construction in navigablewaters and discharge of dredged or fill materials into state waters,including wetlands. COE will be the federal agency to request thiscertification, the timing of which will be tied to the Corps permitapplication review. Public notice for the water quality certification isincluded with the Corps public notice.

5.4.4.2.5 Stormwater Permit

The stormwater permit will be required to address the water qualityconcerns related to any stormwater discharges associated withindustrial activities. This permit requirement implements theregulations set forth by the Environmental Protection Agency (EPA) inSection 301 and Section 402(p) of the Clean Water Act (CWA),primarily contained in 40 CFR 122.26, and is administered by moststates under the delegated authority of the EPA. Time required toobtain a stormwater permit may vary from state to state; however, amaximum of 6 - 8 months is presently estimated.

Development of a stormwater management plan is required for thefacility under this permit, to cover both construction and operation.

5.4.4.2.6 Hazardous Air Pollutants Emission (NESHAP) Permit

This permit, administered by the EPA, covers construction of any newsource of radionuclides or modification to any existing source, underthe National Emission Standards for Hazardous Air pollutants(NESHAP) in Section 112 of the Clean Air Act (CAA). The 1990

amendment of the CAA however indicates that EPA is not required topromulgate standards for radionuclide emissions from a sourcecategory licensed by the NRC, if EPA determines that the NRCregulatory program provides an ample margin of safety to protectpublic health. States retain the right to adopt or enforce standards that

5.4-9

are more stringent than the applicable federal standards. The standardfor DOE facilities is that activities causing radionuclide emissionsshould not result in an effective dose to the public greater than 10millirems per year.

Activities causing such emissions can be the exposure of the primaryreactor coolant to the atmosphere or conditions that result in itsdegassing, and neutron flux in the spaces .adjoining the reactor vessel.Information to be furnished for this permit application includeslocation of the source, technical inputs regarding nature, design,emission estimates, treatment and control-measures to minimizereleases, list of radioactive materials used at the facility and otherpertinent details.

Prior to construction, an initial site study must be performed todetermine the offsite impact by estimating radionuclide emissions

-using an approved EPA procedure. Review and approval process fori this permit generally takes 4 to 6 months.

5.4.4.2.7 Permit for Construction/Operation of Domestic Wells

Construction, modification, and/or expansion of any domestic waterwell are activities that, if undertaken for this facility, will require thispermit. The permit is administered by the State Department of Health,and regulated under the State and Federal Drinking Water Act, andprimary drinking water regulations as applicable.

Any monitoring wells or dewatering wells do not require a permit,although a water well record may be necessary for submittal to theagency, if dewatering wells pump more than 70 gallons per minute(gpm). Also, a statement to the effect that the well will be drilled by astate-certified well driller is required to be furnished. Total timerequired to obtain this permit is approximately 3 months.

5.4.4.2.8 Permit for Construction / Operation of a Public WaterSupply System

As above, the legal authority for the administering of this permit arethe State and Federal Drinking Water Act, and primary drinking waterregulations. Construction, modification, or expansion of a public watersystem as well as its operation are covered by this permit.

For this facility, water supply systems include fire water, domesticwater, demineralized water, cooling water (makeup and circulating),chilled water and heavy water. The permit application will be requiredto furnish complete information regarding general location plans,

5.4-10

surface and/or ground water sources, water treatment plant, waterdistribution systems including improvements and appurtenances,design criteria and calculations.

Approximately 6 months will be required for review and approval ofthis permit to be issued for construction. In addition, operationalapproval must be received following construction, before placing thewater system into operation. The agency issues a written approval after.conducting a final inspection of the completed construction /modification of the water supply system.

!

During operation, chemical and bacteriological self-monitoringrequirements are generally imposed on the facility. A surface water andground water supply operation report form (including water qualityand water production information) is required to be submitted to theagency, at a frequency determined by the nature of the system and asdesignated in the permit.

5.4.4.2.9 Underground Storage Tank (UST) Construction Permit .

Installation of new underground storage tanks which will storeregulated substances (including petroleum) are covered under thispermit, the administering of which is the responsibility of the StateDepartment of Health. The objective is primarily protection ofdrinking water and environmental health and safety. The legalauthority is the Safe Drinking Water Act, and the activity is alsoregulated by the applicable State Underground Storage TankRegulations. The entire permitting process takes 4 - 6 weeks.

5.4.4.2.10 Permit to Construct Sewage Treatment Plant

Construction of an onsite sewage treatment system for this facility, forthe handling, treatment and disposal of sanitary wastewater generatedat the facility rest rooms, showers and dining areas will necessitate thispermit to be acquired. The state department of health andenvironmental control, responsible for the control of water pollutionwill administer this permit. State Pollution Control Act and applicableEPA guidelines will be the regulatory bases for this permitting process.

The application will be required to include physical site description,nature, quantity and characteristics of the waste, treatability of thewaste, details of the treatment system, point of discharge and its impacton the receiving water, and other information relevant to the proposedtreatment method as it relates to NPDES or other permits.

5.4-11

Construction permit for the sewage treatment plant typically takes 4 - 6months for issuance. However, a Permit to Construct will not be issueduntil the NPDES permit (Section 4.2.2) becomes effective. A Permit toOperate must be issued by the agency prior to startup of the sewagetreatment plant at the facility, and the level of operator required will bebased upon the classification received in the Permit to Construct.

5.4.4.2.11 Sanitary Landfill Permit

As the facility generates solid waste, both during construction andoperation, onsite disposal of such waste at a new or modified sanitarylandfill area will necessitate acquiring of this permit. Theadministering authority for this permit will be the State Department ofHealth and Environmental Control, in accordance with the existingState Landfill Regulations. Time requirements for review and approvalis approximately 6 months to a year, depending en the complexity ofthe issueand related factors, such as the public hearing process.

5.4.4.2.12 Radioactive Waste Transport Permit

Transportation of radioactive waste from the facility offsite withinand/)r out of state will require this permit, as well as a 72-houradvance, written notification of such waste shipments. Department ofEnergy (DOE) as the owner of the facility, may have special agreementwith the State Department of Health and Environmental Safety as theadministering agency of this permit, to delineate the conditions forconformance to the permit requirements. The conditions to fulfill suchspecial agreement are considered open at this time. The time requiredfor review and approval of this permit is postulated to be about 3months.

5.4.4.2.13 Hazardous Waste (RCRA Parts A & B)

Storage of any waste, designated as hazardous, for longer than 90 daysonsite, as well as any treatment and/or disposal of such waste insidethe boundary of the facility are activities that will require theacquisition of this permit. Wastes generated at the facility will need tobe characterized at the source, and categorized as either RCRA or non-RCRA wastes. RCRA and/or potential RCRA wastes will be collected inRCRA tanks and the project will have to decide if any of the aboveactivities ( storage over 90 days, treatment, disposal) will take placeonsite to trigger this permit.

The State Department of Health and Environmental Safety responsiblefor solid and hazardous waste management within the state will be theadministering agency for this permit, regulated under the legal

5.4-12

authority of the State Hazardous Waste Management Act andapplicable regulations.

Generally, Parts A and B applications will be reviewed within 60 days.A draft permit (or denial) is subject to a public review and commentperiod of 45 days, when a public hearing may be requested, prior to theDepartment's final decision.

5.4.4.2.14 Notice of Construction / Alteration to FAA

A notification will be required to the Federal Aviation Administration(FAA) for any construction or alteration in the facility at more than 200feet in height above ground level. Existence of cooling towers(especially natural-draft type) and/or stack(s) for exhaust emissions ifany will have the potential to fall in this category.

The notification will be required to include the location of the saidstructure with respect to the nearest city/town and airport(s), its heightand elevation above MSL, and other prominent terrain features in thevicinity. Unless informed otherwise, the notification must besubmitted at least 30 days before the earlier of : a) the start date ofproposed construction/alteration, or b) the date a state constructionpermit is to be filed.

5.4.4.2.15 Radioactive Material Transport (DOT) Permit

This permit will apply to the activity of transporting the plutoniumfrom its storage in different parts of the country to the fuel fabricationcomplex collocated with the ABWR in this facility. However, a specialagreement between DOE and the Department of Transportation (DOT)may be made, to cover this permit requirement by imposing specialconditions unique to this facility for compliance during the operatingphase. In that case, the review and approval time required will beexpected to be minimal. This assumption will be confirmed as moreinformation becomes available regarding this issue.

5.4.4.2.16 Permit to Construct Solid Waste Management System

This permit will be required in conjunction with a Sanitary LandfillPermit to cover construction or modification of any solid wastedisposal unit onsite. The State Solid Waste Management Regulationswill govern the requirements of this permitting process, and theadministering authority for this permit will be the State Department ofHealth and Environmental Control.

5.4-13

In addition to the general description, location, zoning, process flowetc., and type, quality and quantity of solid waste generated in thefacility, additional information to fulfill the application requirementsfor this permit include names, addresses and telephone numbers of thedesign engineer and the company official directly responsible for theplant Solid Waste Management System.

Time for review and approval of this permit may vary, from 4 monthsto a year, depending on the complexity of the system and the outcomeof public hearings.

5.4-14

GREENFIELD SITE =I

ISITING LAND FIELD DATA I:::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::Ao, ::::::::::::::::::::::::::::

iI I

_ iiiiii i iii iiii i i i iiii

PROGRAMATIC EIS I SITE SPECIFIC EIS III --- -- L--._,li il

CONS'IttUC'IION

YEARSl I I i I I I

0 1 2 3 4 5 6

DOE SITE

iii iiii

SITE SPECIFIC EIS

i ii

PROGRAMATIC EIS

, r, ,l_ll, _ o. =..., i.., ... °. o ,. , .

,.'0. " ".'., ' .0" ' " ,o' °.' ' ".. ".'o'.'.' • '.' ".. " • "." °0' " •

III

CONSTRUCTION

FIGURE 5.4-1 This graphic presents two optimized permitting schedules for the ABWR

complex located at a "greenfield" site and located at an existing DOE facility such as theSavannah River Site. The view shown here is "optimized" to the extent that some sequential

tasks are conducted in parallel in order to save time. The durations of the individual tasks arepreliminary estimates based on agency requirements and past experience with similaractivities. Both scenarios assume that all activities "go according to plan" and the durationsrepresent the minim_m time to free the project for construction. In the case of the greenfield

site, the various public relations and awareness programs conducted in support of the project areassumed to be successful to the extent that public opposition is not a major factor in scheduling.

5.4-15

2 - 1 0 1 2 3 4 YEARS

I J I I ! IPROGRAMATICEIS _,,?--:_:"'-::"'_:":/-:--'"'"'_-'/_:h:h:::::l Ra3

ISITING STUDIES & SELECTION :::::::::::::::':':! i

LANDACQUISITION i:: ;: ;: :: :: ::-:: :: ;:lI

FIELD STUDIES t:-:--_'-;--'-'_:'_-'-:i-:;..ai I":.:.':.:::.":.":.:::.:.::::--':'":-".:-":':.':'":'":'-':':::-":':J

SITE SPECIFIC EIS iII

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PUBLIC WATER SUPPLY PERM I t v I_PJ_T _:.::.: .-:u, :pRE:vAi ; 6i:'IUSTCONSTRUCTIONPERM .::::::::::::::::::::::_

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SOUD WASTE MGM. SYS. PERM. I F].[.S.[.[.[.S.S.].S.].]t::_.::_.::..::_.::_.::_.::_.::..::.i_.i .i

FIGURE 5.4-2 Preliminary environmental permit schedule for an ABWR greenfield site. The major environmental permitsthat could apply to an ABWR site are listed above with time lines determined for the average case. This presentation assumesthat adverse public involvement in the EIS/permit process (which could stop the process altogether) can be kept to a minimum.The permit time lines shown in the diagram start at the time the permit is submitted to the responsible agency. Preparation ofall the permits for submittal is assumed to start about year 3 - 3.5 during the field studies program. A similar list of permits

,ould also apply for a DOE Facility site but the permit proce: be simplified because existing site Permits could beto accommodate the ABWR.

0 1 2 3 4 5 6 YEARS

, i -! I i I IPROGRAMATICEIS :_.:-"_:._.'.._.'.'-..:---_--_.._-.":--.:._.:."---:._:_::_:1RI3D

iSITING STUDIES& SELECTION _ i

SITE SPECIFICEIS ,-::.-::.-::.-::.-::...;:.-::.-i:.-::.-::._-::.-::.-::.-::.-::.-::.-::.-::.lR00I

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.......................

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FAANOTIFICATION :::::::::::::::::::::::::::::::::::::::::::::::I

SOUD WASTE MGM. SYS. PERM. ==========================================================II

FIGURE 5.4-3 Preliminary environmental permit schedule for an ABWR at a DOE site. Compared to a greenfield site (Figure5.4-2), considerable schedule compression is possible at an existing DOE facility such as the SRS. This graphic shows the mostfavorable case for the ABWR complex located at a DOE site. The rational behind a slightly over 2 year schedule is the largebody of engineering and safety information available for the ABWR coupled with essentially all the information required forenvironmental assessment and permitting is in place at a DOE site. Instead of new permits, a DOE site would involve ,medification of existing site-wide permits and all background information necessary for the environmental assessments hasalready been collected and analyzed.

6.0 DEVELOPMENT REQUIREMENTS

6.1 Development Requirements Overview:

In summarizing the development requirements in Phase 1A of this study, it was concluded

that the proposed means of dispositioning weapons plutonium by conversion to mixed

oxide and its use as fuel in an ABWR entailed no new technology and needed only

verification or validation of already existing technology that had not been used in this

country for some time. Detailed evaluations conducted during Phase 1B and 1C have only

served to confirm this position. Specifically, it has been found that:

• No system level changes are required for the ABWR to utilize MOX fuel;

• Technology to fabricate MOX fuel from Plutonium oxide is well understood although

such fabrication has not been carried out in this country in almost two decades;

• The technology that is being implemented to fabricate MOX fuel from reprocessed fuel

being implemented in other parts of the world is readily adapted to making MOX fuel

from weapons Pu while requiring less shielding and is potentially easier to handle.

By the same token, unlike reprocessed plutonium, weapons plutonium is potentially moreattractive for diversion.

Limited development/engineering verification tasks were outlined in the Phase 1A and

Phase 1B reports in the following areas:

1. Safeguards:

Disposition Process Simulation, Literature Survey

Software Development for Safeguards C&I Integration

2. Fuel Cycle

Validation of Nuclear Methods by Monte-Carlo Techniques and Benchmark Data

Confirmatory Testing for MOX Fuel Fabrication (with and without Gd)

Lead MOX Fuel Pin Tests

3. Tritium Production

6.1 - 1

No new areas requiring development or engineering verification were identified during this

phase of the study. Further detailed evaluations however were conducted in specific areas

during this Phase of the study. These include the specific development tasks needed for

lead fuel tests (given in Section 2 of this report), long-lead engineering tasks associated

with the MOX fuel fabrication plant and development tasks for tritium production. No

development tasks were found needed for the reactor system.

6.2 Development Requirements for MOX Factory

The factory for fabrication of MOX fuel is on the critical path to the disposition

process. As pointed out earlier, a certain level of safeguards implementation would have

taken place by the time the pits are converted into MOX pellets as it would take several

weeks to convert this material back to a weapons usable form. For this reason, it is

desirable to implement MOX fabrication as quickly as possible.

The technology for making MOX fuel is already being implemented in foreign

countries. Evaluations have been conducted on adapting and ! lproving this technology for

weapons Pu disposition. One area which has been identified for further evaluation

involves the surveillance equipment used for safeguards and security control. Unh_:e

plutonium from reprocessed fuel which is highly 7 active, weapons plutonium has a low

level of 7 activity. For this reason, material accountability or surveillance cannot be

dependent upon instrumentation based on _, activity per se. Alternative, positive

interrogation measures have to be implemented, including the use of Cf scanners. While

such scanners have been used in the past, their use has principally been for fully assembled

fuel rods. Their application to on-line measurements is expected to require additional

development and engineering verification. Similarly magnetic sensors have been used for

identifying Gd bearing rods and once again, some additional development and engineering

validation for application in the proposed MOX factory will be needed.

Recent advancements in both laser and fiber optic technology have been

incorporated into real time measurement systems for nuclear material. Although still in the

development phases, this technology appears to be directly applicable to the accountability

and certification measurements required during fabrication of MOX fuel. These systems

are particularly attractive because all of the electronics are located outside of the

contaminated area and fiber optics are utilized to transfer signals for evaluation. Current

development by national laboratory personnel has resulted in prototype measurement

systems for metal and isotopic compositions. Since these types of systems provide

measurement results in minutes instead of hours or even days associated with analytical

chemistry measurements utilized for MOX fuel fabrication, process flow rates can be

6.2- 1

improved and in process storage requirements reduced. These modifications in the MOX

fuel fabrication techniques are expected to • _ult in a more cost effective facility and

reduced operating costs.

Resources for preliminary engineering required for this task are estimated to be 4

engineers for a period of 9 months.

6.2 - 2

6.3 DEVELOPMENTPLANSFORTRITIUM PRODUCTIONi

6.3.1 Background

As part of the close-out of the Tritium Target Development Project (TTDP) plans

were prepared for completing development of the tritium target to support a wide

range of deployment options (Apley 1992) (Reference _1). .The development phase

of this completion plan consisted of the following six areas"

• Evaluation of deployment options

• Target rod design

• Target rod fabrication• Tritium extraction

• Handling, storage,transportationand waste characterization

• Preparationof a target rod qualificationpackage

• Lead test assembly(s)

The basic premiseunderlyingthe TTDP completionplan was that the initialphase

would be completed prior to selection of a new tritium production goal and

deploymentoption. As a result, it includedtasks such as preliminarycore and

fuel cycle design, safety assessments,evaluation of changes required to the

plant and its operatingand safety documentation,cost and scheduleestimates,

etc. for a range of plant and core designs.

However,this preliminarydesign and analysiswork for tritiumproductionin the

ABWR has been incorporatedin the PlutoniumDispositionStudy. Further, the

final core design, safety analysis, etc. needed for tritium production is

includedas part of the ABWR completioncost. Thus,the developmentrequiredfor

tritium production in the ABWR is less than that envisioned in the TTDP

completionplan.

Another important difference relates to the need for Lead Test Assemblies

(LTA's). The targetdevelopmentprogramfor the NPRwas originallyplannedsuch

that the requiredlevel of confidencein targetperformancewould be providedby

theATR tests plus the designand vendorqualificationprograms. In thiscontext

6.3-I

LTA's were considered as a tool to evaluate extended burnup oN other performance

enhancements during operation of the NPR but not as a prerequisite to startup.

The final TTDP completion plan, however, was structuredto support a broader

rangeof optionsincludingan off-the-shelfcapabilitythat could be implemented

on short noticein existingLWR's. Itwas this goal that led to recommendingone

or more LTA's to supportthe range of potentialdeployment scenarios.

The ATR tests demonstratedthat the getter-barriertargetdesignmet or exceeded

all of its designrequirements. Further,the targetrodsoperateundermuch more

benignconditionsthan the fuel rods. For example,post-irradiationexamination

of the target rod (Lanning1992) (Reference2) confirmedthat its claddingdid

not experienceany chemicalattack or mechanical interactionwith its internal

componentsand is essentiallya free-standinggas-pressurizedtube.

As discussedlater it is plannedto modifythe standardABWR fuel bundlehardware

to facilitate remote installationand removal of -3,500 target rods a year.

Under currentNRC practiceLead Use Assemblies(LUA's)are irradiatedto confirm

the performanceof fuel assemblyhardwaremodifications. LUA's do not normally

addressthe rod performanceotherthan mechanicalconsiderationssucha vibration

and wear. However, since prototypic target rods would be available from the

large lot fabricationdemonstrationit is proposedthatthey be includedin these

LUA's. This will provideaddedconfidencein the overalltarget rod performance

plus a final validationof the fabricationqualificationprogram.

LTA's should be considered for inclusion in the ABWR at startup to explore

extendedtarget burnup capability. While we believethe getter-barriertarget

design is capable of much higher burnups, irradiationof the target rods for

multiple cycles is not necessaryto meet the tritium production requirements.

As a result,extendedburnup testingof target rods has not been costed as part

of the developmentprogram.

6.3-2

6.3.2 Target Development Tasks

The following sections provide a brief description of the applicable developmenttasks and their estimated costs based on the information available at the

completion of the TTDPin early 1992. The costs have escalated to January, 1994

at 4%/yr.

{ommercia! LiAI02Pellet Fabricationpemonstr_tion- $O._M

The purposeof this task is to optimizethe pellet fabricationprocess,qualify

vendors and finalizethe pellet procurementspecification.

Provide a Benchmarked,ProductionNDE System- $I.1M

The TTDP successfullydevelopeda prototypeNDE system for verificationof the

quality of the barriercoating on the target rod cladding. This task provides

a production-scaleNDE unit for use in the large-lotfabricationdemonstration.

It is presumed that the same unit, if successful,would be used in production.

Larqe Lot Tarqet Rod FabricationDemonstration- $2.7M

Fabrication of approximately I00 complete, full size target rods, including

barrier-coatedcladding,getters,pellets,and liners. These rods would be used

to validatethe componentfabricationand rod assemblyprocesses,qualifyvendors

and finalizethe facilitylayoutsand procurementspecificationsfor full scale

production.

CompleteThe D2Z.I2 Correlation- $2.7M

Through a comprehensive series of ex-reactor experiments validate that

deuterium/tritium(D2/T2)correlationscan be reliably used to predict the

effects of time, temperatureand pressureon the permeationof tritiumthrough

aluminized barriers and the extraction from and adsorption in the target rod

materials. This data is needed to finalizethe mod31s for target rod behavior

6.3-3

during normal in-reactor operation and transient conditions and during

extraction.

Validate the Pello1;/Getter/Barrler Kinetics Model - $2,7M

Conduct the ex-reactor experiments neededto validate the kinetics model and the

performance of the integrated target rod system.

Full-lenqthD2 ExtractionTests - $3.)M

Validation of existing extraction models and assumption would be achieved by

constructing a prototypic length extraction furnace for testing target rods with

deuterium-loaded components, The results of these tests would be used to

finalize the design of the production-scale extraction furnace.

Fqll-length Breach Tests- $0.6M

Conductex-reactortransientbreachtestingof full-sizetarget rods to confirm

the safetymargin betweenexpectedperformanceduring in-reactortransientsand

conditionswhere target rod claddingfailurecould result in ejectionof LiAl02

from the rod.

Fue.l/TarqetBundleMechanicalDesiqn and TestincI - $6M

The standard ABWR fuel bundle hardware will be modified to facilitate the

installationand removalof -3,500targetrods ayear. Since the targetand fuel

rods have identicalexternaldimensionsand the ABWR fuel bundlewas originally

designed for remote reconstitutionno difficultiesare anticipated. However,

based on currentNRC practiceex-reactorflow, vibrationand mechanicaltesting

of the modified fuel bundle hardware and irradiationof six LUA's is proposed.

The LUA's would containstandardBWR uraniafuel plus target rods from the large

lot fabricationdemonstrationand would be irradiatedin existing BWR's. The

costs for this task include design and ex-reactor testing of the modified

6.3-4

subassembly hardware, fabrication of the urania fuel rods and the necessary

safety and licensing documentation.

AutomatedBundle DisassemblyHardwar@- $0.5M

The ABWR fuel bundle was designed to be disassembled remotely using manual

tooling. However,this operationwas intendedfor occasionaluse rather than a

high volume productionoperation. Experienceindicatesthat with the current

manual tooling a single crew can process four bundles per shift. While this

would be adequate using multiple crews, it is proposed to develop an semi-

automated disassembly machine to minimize worker exposure, provide higher

throughputsand minimizeoperatingcosts and the potentialfor errorsor hardware

damage.

ProjectManaqement and Restart- $2.5M

It is assumed that the target developmentprogramwould be performedby a DOE

laboratoryteam such as the PNL-Westinghouseteam that conductedthe TTDP. This

task coversthe programrestartand managementcosts for completionof the target

developmentfor the ABWR.

Compile the Target Rod QualificationPackaqe- $2.5M

This is the final deliverablefrom the target developmentprogram and provides

the basis for final design of the target rod, the fabricationand extraction

facilities,procurementspecificationsand safety review/licensingsubmittals.

It includesthe resultsof testingand analysisused to validate the target rod

design and its performance models, the basis for and results of vendor

qualification demonstrationsplus the information necessary to support the

handling, storage, transportationand waste characterizationassociated with

tritiumproduction in the ABWR.

6.3.3 Other Considerations

6.3-5

In order to extend the burnup capability of the target rod it wtll be necessary

to perform adequate non-destructive and destructive examinations of irradiated

full-length rods. Based on the post Irradiation examination results from the

TTDP (Lanntng 1992) the key examination capabilities would be:

• Visual and dimensional Inspection

• Sipping to check for cladding leaks

• _ " Neutron radiography to verify the mechanical Integrity and location of the

LiAI02 pellets

• Rod puncture and plenum gas analysis

• Rod cut-up, metallographyof cladding and internal rod components and

measurement of Li6 depletion and tritium retention in the LiAlO2 and

getters.

The first two items are fairly standardexaminationcapabilitieswhich can be

performedat a number of locationsincludinga reactorspent fuel pool. Neutron

radiographyfor the TTDP was performedin the HFEF-Northfacilityat DOE's ANL-

West site. While the TTDP target rods were only 4-feet long the HFEF-North

neutronradiographywas setup to handle TREAT loopswhich are much longer. Thus

it is conceivablethat a full length ABWR target rod could be radiographedat

HFEF-Northusing multipleexposures.

All of the destructiveexaminationsfor the TTDP were performedin the PNL hot

cells at Hanford. The in-celltritiumenclosuresnecessaryfor this work were

sized for 4-footlong rods and have been removed. However,both the PNL and ANL-

West hot cells are large enoughto installthe in-celltritiumenclosuresneeded

for destructiveexaminationof full lengthABWR target rods. Based on the TTDP

experienceit is expectedthat this capabilityfor full length ABWR target rods

could be establishedin either of these facilitiesfor about $5 million. This

presumesthat existingmetallographicand analyticalfacilitiesfor radioactive

samplescould be used which was the case for the TTDP.

6.3-6

Whtle the above target rod examinations are not unique to the ABWRor essential

to meeting the tritium production requirements, experience Indicates that if

tritium production is a long term mission it is prudent to plan for these typesof examinations.

6.3.4 References

1. Apley 1992: PNL-8142, "Tritium Target Development Project Executive

SummaryTopical Report" - Appendix G, September 1992.

2. Lanntng 1992: PNL-8133, "Final Report on the WC-1 LWRTarget Rod

Irradiation Test and Post Irradiation Examinations (Task 3), July 1992.

6.3-7

7.1 SAFEGUARDS REQUIREMENTS FOR PuAND TRITIUM TRANSPORT

Introduction and SummaEy

The previous report in this series provided a framework for assessingthe impacts of facility siting and processing technology on thesafeguards and security of various options for Pu disposition byirradiation in an ABWR. This framework actually must cover more thanthe Pu Disposition Complex alone. The evaluation must also encompassthe critical interfaces with the Nuclear Weapons ComplexReconfiguration Program for the Pu feed material used in MOx

fabrica£ion and for tritium production and recovery (if needed), inorder to provide a complete assessment of the safeguards and securityissues. The present report continues the evaluation by furtheranalyzing the safeguards and security of the siting options for thereactor and fuel cycle facilities with the objective of narrowing theoptions presented in the earlier report. Using reasonableassumptions it was possible to narrow the list of siting options fromsix to four.

The safeguards & security issues involved in the tritium productionoption are also evaluated. If the configuration of the target rodsin the production core remains unclassified, it is expected thatirradiated target rods could be shipped in LWR spent fuel casks withthe level of safeguards and security normally associated with spentfuel shipments today. If the configuration is classified, an LWRshipping cask could still be used but additional escort requirementswould apply. Shipping tritium product from an on-reactor-site tritiumextraction facility to the Savannah River Plant does not, in and ofitself, require the use of an SST. However, given the strategic andeconomic value of tritium in a single shipment and the relatively fewshipments needed, use of an SST is considered prudent.

7.1-1

Sitinu of Fuel Cvale Facilities and the Reactor

The previous report in this series (Ref.l) provided a initialframework for assessing the impacts of facility siting and processingtechnology on the safeguards and security of Pu disposition byirradiation in an ABWR. In this report analysis of the safeguardsand security (S & S) impacts of the various siting options outlinedin Ref.1, and repeated in Fig.1 herein for convenience, is continued.

In the earlier report it was found possible to express a Figure ofMerit (FOM) for the S & S assessment as:

FOM = f(CD, DP, TR, SE, t) (i)

where CD is the dispersion of the complex (i.e. the number of sitesinvolved), DP is the diversion potential of the material beingprocessed during the Pu disposition fuel cycle, TR is thetransportation risk of a single intersite transfer, SE is the SystemEffectiveness of the anti-diversion measures (probability of alarm,proper situational assessment, and neutralization) at any point inthe process and t is the time required to process the entire i00 Mg(megagrams) of excess weapons plutonium. The complex is not yetsufficiently defined to quantitatively evaluate figures of merit.However, several of the variables permit intuitive assessment. Forexample, the longer the time required to process the entire availableinventory, the greater the diversion risk. Similarly, it may bepossible to intuitively evaluate the siting options presented in theprevious report and down-select from the six theoretically possibleoptions to some smaller number. This logic is demonstrated below.

In Figure 1 the Pu flow for all options starts with WeaponDismantlement and we assume; for this report, that this operationwill always take place at the pantex Plant in the interests ofnational security. The final location in the complex is the reactorsite, which we shall call Site A. Thus the Pu flow is always fromPantex to Site A, possibly with intermediate stops. Additionally, weassume that the Pantex site is not acceptable as the site of the Pudisposition reactor (Site A), either because of limited water supply,site size and surrounding population, or the potential requirementsfor transparency in verification of pu disposition and the consequentpreuence of foreign nationals on-site. These assumptions eliminateoption 1 from further consideration.

The five remaining options can be further reduced if we invoke otherarguable, yet reasonable assumptions. We assume that transportationrisks, based on safety, safeguards, or cost are excessive if threesuccessive shipments of Category I Pu are required to deliver eachbatch of fabricated pu fuel to the reactor and that this

transportation continues over the program lifetime. Option 6 canthen be eliminated from consideration.

7.1-2

IOom.o"_ne_,-' ' | C_v._o_ !,.ea s1 ._IM.t,i Mox..r. ___ Fuel Fueled,, : ,, ___ Fabrk:atk3n OYI'ION I

SINGLE SITE

Weapons I@A3X MOX

Dismantlement Conversion Fuel Fueled OPTION Z2 SITES

Metal MOX j MOX

Conversion Fuel ,, _ ! FueledF_:_l:ion ALWR OPTION $

2 SITES

Metal MOX

Dismantlemen Conversion Fuel FueledALW_ OPTION 5$ SITES

MOX MOXMetal Fuel FueledConversion Fabdca_on ALWR OPTION 6

4 SITES

Yucca Mountain

Reposltorg I I

FIG.I FACILITY/LOCATION OPTIONS FORABWR DISPOSAL OF WEAPONS PLUTONIUM

Four options remain; three options with a single Pu shipment in thefuel cycle (options 2,3,4) and a single option (#5) with two pushipments in the fuel cycle. These options are shown in Figure 1 ofthis report and detailed as follows:

One shipment :

° (2) Metal conversion and MOx fabrication at PANTEX; reactor atSite A

• (3) Metal conversion at PANTEX; MOx fabrication and reactor atSite A

• (4) Metal conversion, MOx fabrication and reactor at Site A

Two sh iDment s_• (5) Metal conversion and MOx fabrication at Site B (SRS, RL,

LANL);reactor at Site A.

T1_ere are no other options for the fuel cycle facilities exceptforeign sites or (presumably) "politically difficult" sites likeRocky Flats or LLNL.

From a safeguards and security standpoint, this analysis demonstrates; the importance of the selection of Site A for the reactor. Pantex and

Site A are the end points for the pre-irradiation fuel cycle anddefine the shipping lanes for plutonium (Pantex m Site A) and spentfuel shipments (presumably from Site A I Yucca Mountain). Candidatesites for the metal conversion and fuel fabrication steps might beeliminated because of large shipping distances, or be enhanced by theexistence of easily modified or upgradable existing facilities.

The importance of shipping distances and locations for Site A and the

fuel cycle facilities (Site B) is emphasized by examining separately,the safeguards requirements for transport of feed material andcompleted MOx subassemblies. Shipments of Pu feed material for MOxfabrication do not appear to present a logistics problem other thanthe requirement for Safe Secure Transport (SST). Feed materialshipments will be classified either Category IB or IC SNM dependingupon the shipment form (metal is IB, PuO2 is IC) as shown in Figure2, reproduced from DOE Order 5633.3A (Ref.2). Intersite shipmentswill be made under the auspices of the Transportation SafeguardsDivision of the Albuquerque Operations Office (SST); intrasiteshipments will escorted by field element couriers or contractorsecurity forces as specified in Site Master Safeguards and SecurityAgreements required by Order 5632.2A (Ref.3).

If the fuel cycle facilities and the reactor are co-located,shipments of MOx subassemblies from the fabrication facility to thereactor may be made under the supervision of site protective forces;SST is not required. If the reactor and the fuel cycle facilitiesare not co-located, S/A shipments will be treated as Category IID SNM(Ref.4) and SST is required by DOE Order 5632.2A (Ref.3).

7.1-4

,, i iii

Attrac- FU/U-23J I ContnLued U-2J5 Ilemllr_ _ gtLveneus Category Category J_ntorielo --rq

Level I IX XZX xvl I II XII ][114 Ca t;_or]r _m(Uuawrxzxea xN xo8) (gUA_IZIE8 XNX|) _t_t t | io

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, i , , i i i i i

fUflZ P RflDIICT._Pits, major compeflefll:l,huttoflo, Jnqotn _2 >-0.4<2 _O.2<O.4 <0.2 _S _l<S _O.4<1 <0.4re<notable metal, ndirect ly convertiblematerials

t •

III O ll-ORltOl5 NRTEII I AI, B

Carbides, oxides,solutions (>_ 25g/1),nit<area, el<., I[uoi C 2:6 ]_.2<6 >--0.4<2 <0.4 >-20 >i5<20 :_2<6 <2elements and annembl/ene

-4 alXoye and mixtures, UF,

or ur£._(_> 501 enriched|| , ii • i i i i

_nIJ0W ORADE 14JTZRI/UG9

Solutions (1 to 3S9/1 )pro<can rneldtlearoqp, LrLng extensive D H/A >16 >3<16 <3 H/A _SO >_8<50 <8reprocenaLngt moderate|yirradiated material, FUn,(except _aote), UF, or UF_

..J.,_Ot < SOt enrich.dL_._ - .....

_ _m mXZnZALS N/A Nix Jqla * n/A n/a nlA *llLghIy I.clcadlLlted rares,lOllltlon8 I-_ I g/l), Eu:an/um containing e Reportable quantit:lLoo e_o • Reportable quontLklto ere lepmrlLdblo< 201k U-235 |any form any Cat_ego[y I1/ rtgilcd3Loel at Q_mt|tteequanr.[r.y) Cat::egow ZY regmJrdZeaJg of amount.

amount.

1 The lower limit tar Category IV is equal to reportable quantities in this Order.

2 See paragraphs 3b end 3c for HC&/trequirements for tritium and depleted uranium.Bee

!

Fig. 3 Nuolear Haterlal Bafeguarda OategorLes "0

Subassembly shipment from Site B to Site A might raise alternativerisk issues. Each subassembly (S/A) will _eigh about 200 Kg. in thecrated shipping condition, with each shipping _rate holding two S/As.Since SST trailer loadings are limited to 10,0u_ _.(~4500 Kg.), 14SST shipments/year would be required to provide annual core loadings(324 S/A) for the two-reactor case. Commercial carriers, presumablywith larger load limits, might reduce shipment risk (directlyproportional to the number of shipments) even if escort vehicles wererequired.

It is not possible to further reduce the viable siting options usingS & S criteria alone, although option #5 may be le_ desirable sinceit involves two Pu shipments (one Category IB and one Category IID)in the fuel cycle. It does, however, provide DOE more flexibility inSiting of thereactor and the fuel. cycle facilities which arenowindependent.

This analysis has been carried out in accordance with the specificrequirements of current DOE Orders (5632.2A and 5633.3A) governingprotection, control and accountability of SNM. But current Orders donot specifically address the unique circumstances presented by theuse of Pu in ABWRs. Both Orders 5632.2A and 5633.3A provide for thegranting of exceptions in accordance with Order 5633.2 (Ref.5) suchthat more flexibility could be introduced into the SST requirementfor the shipment of MOx subassemblies. Also, it is difficult toascertain how these Orders might be modified in the future and howthey may be impacted by international safeguards requirements and thepossible needs for transparency and bilateral symmetry in Pudisposition (See Sect.7.4).

7.1-6

Safeauards & Security Reauirements for the Tritium

Production ODtion.

A preliminary evaluation has been made of the safeguards and securityissues involved with the production of tritium in an ABWR. Asindicated by footnote 2 in Fig.2 (and paragraph 3b of DOE Order5633.3A), tritium is classified as a "nuclear material of strategicimportance" although it is not a "special nuclear material"° However,as a very costly material, tritium accountability is required tohundredths of a gram (I00 Ci) (op.cit.). Quantities of tritium inexcess of 50g. are treated as equivalent to Category III specialnuclear material (SNM); all other "reportable" quantities of tritiumare Category IV.

An earlier report of this series (Ref.6) showed that, if needed,tritium could be produced in goal quantities in an ABWR using auranium-, rather than Pu-, fueled core. Table 6.3-1 of Ref.6 showsthat each tritium target rod will contain 11.4 KCi (~1.2 g) oftritium at end of a reactor cycle. Consequently a single rodconstitutes a Category IV material, while the shipment of - 1/3 ofthe annual production from an ABWR core (~1.4 Kg) must be treated asa Category III material. The target rod, having a radioactivitylevel in excess of I00 R/h at a distance of 1 meter is considered asAttractiveness Level E.

The reference concept for the tritium production option assumes thattritium target rods will be shipped from the reactor site (Site Aherein) to the Savannah River Site (SRS) for tritium extraction,purification, and storage. Based on current DOE guidance certainaspects of the target rod fabrication, irradiation, and tritiumextraction for an actual production operation would be classified butthe general configuration of the target rod would be unclassified.Thus, for purpose of this safeguards and security evaluation, it isassumed that shipments of target rods would not be considered asinvolving "classified configurations".

An alternate tritium production concept calls for co-location of thetritium extraction facility with the reactor. In this latter case,tritium gas, probably absorbed on uranium metal ("hydride beds"),would be the shipment form for tritium. Such shipments would notinvolve classified materials.

In either of the above cases, each shipment could be as large as 1/3of the annual ABWR production of -4 Kg.(4 x 107 Ci) (Ref.6). Ashipment of that size (-1.4 Kg.) has a value in excess of $20 millionand, on this basis, certainly warrants special protective measures.Irradiated tritium target rods are highly radioactive, requiringshipment in a spent fuel cask and are, therefore, inherentlysafeguarded; extracted and purified tritium gas in approved shippingcontainers is predominantly a low-level beta- emitter and not self-safeguarded.

DOE Order 5632.2A (Ref.4) provides the following baselines protectionrequirements (paraphrased) for Category III and IV quantities of SNMin transit:

7.1-7

(i) Domestic shipments of classified configurations ofCategory III quantities of SNM shall be made by Safe SecureTransport (SST) approved by the Albuquerque OperationsOffice.

(2) Domestic shipments of unclassified Category IIIquantities of SNM may be transported as specified in (I)above, "as deemed prudent and appropriate, byagreements between the Manager, Albuquerque OperationsOffice and the respective Heads of Field Elements."

(3) "Packages shall be sealed with tamper-indicating seals."

(4) Domestic shipments of unclassified Category IIIquantities of SNM not transported by SST.may be shippedeither by government-owned truck or by exclusive usecommercial carrier, or by rail. At least two escorts, atleast one with a "Q" clearance and the other with an "L"clearance must accompany the shipment.

(5) Domestic shipments of Category IV quantities of SNM may beshipped by either SST, or commercial carrier providedproper shipment traceability and package dispatch andreceipt requirements are met.

We can conclude from DOE Order 5632.2A that tritium shipments fromthe reactor to the Savannah River Site will require Category IIIsafeguards and security unless the annual production is subdividedinto unreasonably small quantities. The following restraints willlikely apply :

(i) If target rod configuration remains unclassified, or evenif the configuration is classified, target rods couldprobably be shipped in a spent fuel cask since theradiation level would make them self-protective. Anexemption from shipment by SST may be required.

(2) If tritium gas or hydride beds are shipped, SSTmay be required. As a minimum, cleared escorts will berequired.

Finally, enriched lithium-6 feed material and unirradiated targetrods (if the configuration is unclassified) are treated as CategoryIV S_4 and may be shipped by commercial carrier without escort, butwith proper notification, shipment traceability, and packagereceipts.

7.1-8

7.4 APPLICATION OF SAFEGUARDS AND SECURITY CONCEPTS

TO THE ABWR IN A PLUTONIUM DISPOSITION MODE

This evaluation was based on full compliance with existing DOE ordersrelated to safeguards and security of SNM. However, these orders areexpected to continue to evolve and they do provide for exemptionswhere appropriate. While the existing DOE Orders provide a broadframework covering all aspects of SNM protection involved in thePlutonium Disposition program, they were not developed tospecifically address some of the unique features of this program.These include the transportation of large numbers of mixed oxide fuelassemblies containing a few percent plutonium and the transportationof irradiated tritium target elements.

Other factors which are likely to influence the safeguards andsecurity plan for a US. Plutonium Disposition program include:

• Bilateral or international agreements related to the disposalof excess weapons plutonium,

• The need for reciprocity and symmetry in implementation ofthese agreements.

• Other applicable experience and standards such as the IAEASafeguards system, and

• The need for prudent practice to assure the public that thesafeguards and security measures applied are adequate.

In this context we believe it is appropriate to plan for fullimplementation of existing DOE Orders is this area but continue toexplore alternatives that may be less expensive yet provide adequateprotection.

7.1-9

7.5 Reference8

i. GE Nuclear Energy, "Study of Pu Consumption in Advanced LightWater Reactors; Evaluation of GE Advanced Boiling Water ReactorPlants, Compilation of Phase IB Task Reports", NEDO-32293, RFPDE-AC03-93SFI9681, September 15, 1993.

2. DOE Order 5633.3A, "Control and Accountability of NuclearMaterials", February 12, 1993.

3. DOE Order 5632.2A, "Physical Protection of Special NuclearMaterial and Vital Equipment", January 17, 1989.

4. USDOE, Office of Safeguards & Security, "Guide to DOE Order5633.3, Control and Accountability of Nuclear Materials, DraftGuidance", April 1990.

5. DOE Order 5633.2, "Control and Accountability of NuclearMaterials: Responsibilities and Authorities", January 29, 1988.

6. GE Nuclear Energy, "Study of Pu Consumption in Advanced LightWater Reactors", Evaluation of GE Advanced Boiling Water ReactorPlants", NEDO-32292, RFP DE-AC03-93SFI9681, May 13, 1993.

7.l-lO

8.1 Cost and Schedule

II 8.1.1 IntroductionI

The Phase 1C tasks concerning project costs and schedules included the following:

• Review and update of the structuresaccount of the Phase 1A baseline ABWR capitalcost estimate.

• Development of pre-conceptual cost estimates & schedules which address the fourprimary Phase 1C alternative cases being studied, which are: (1) disposition of 100MT of Pu in 25 years, (2) disposition of 50 MT 1% in 25 years, (3) disposition of100 MT within the ABWR license of 40 years and reactor lifetime of 60 years, (4)disposition of 50 MT Pu in 40 and 60 years.

• Development of a typical cost comparison matrix to compare costs of the projectassuming a "Greenfield" site with an operating DOE site.

• Development of preliminary transportation costs for transporting plutonium feedmaterial, new and used fuel assemblies, and plutonium waste materials to the 1%Disposition complex or to appropriate off-site-facilities.

8.1.2 Data From Available Existing Cost Studies:

Account number 21 (Structures and Improvements) of the General Electric capital cost estimate

for the ABWR reactor complex which was used as the basis for reactor capital costs for the Phase

1A is being reviewed. Information which had been developed for recent ABWR commercial

proposals, cost studies for the GE Simplified Boiling Reactor (SBWR), and the quantities

developed from the construction of the ABWRs in Japan is being used as the basis for the review.

Appropriate cost updates being made use the DOE cost guidelines as outlined in the "Cost

Estimate Guidelines for Advanced Nuclear Power Technologies," ORNL/TM-1007.

The ABWR capital costs will be updated using a cost database that is consistent with the

information being developed for the ABWR First-of-a-Kind Engineering (FOAKE) program and

other programs. The following sources of data are being utilized to update the capital costestimate for the ABWR:

8.1-1

a. Refinement of bulk commodities initially developedfrom the construction of the ABWRsin Japan (K-6/7).

b. Cost data which had been deveioped for a recentABWR commercial proposal.

c. Recently completed detailedcapital cost estimate for GE Simplified Boiling WaterReactor(SBWR) using the DOE cost guidelines as outlined in the "CostEstimate Guidelines forAdvanced Nuclear Power Technologies"(ORNL/TM-10071/P.,3)andBechtel's experiencein constructing both fossil and nuclearplants.

d. Historical data on field installation costs for equipment and materials where appropriate.

The updatedcapital cost estimate for the ABWR will be based upon the above mentioned DOE

guidelines. The level of detail is a function of resourcesandtime. Forexample,as an integralpart

of the work performedfor FOAKE,a detailedcost estimate is plannedfor completion in

June/Julyof 1994. An earlier completion date of April/May 1994 to supportthe Pu disposition

study is feasiblewith less details. All cost estimates will be presentedin the EnergyEconomic

Data Base (EEDB) format. Additional interactionwith DOE and ORNLis anticipatedin orderto

assureproperapplicationof the guidelines to the new cost data. These activities areexpected toprovidethe most creditable cost estimate to date for an ALWR.

8.1.3 Development of Phase IC Alternatives Cost Estimates and Schedules:

Capital costs for the four basic alternativecases defined above for Phase 1C are being developed.These estimates will use the revised GE ABWR one-reactor estimate discussed above as the

baseline case. Estimates for the other multiple reactor cases will then be factored from thisbaseline estimate.

Project schedules for the four Phase 1Ccases areshown in Figures 8-1 through 8-5.

Cost cash flows for the updated costs over the project duration for each of the four cases will

also be reported.

8.1-2

The approach to revenue calculations is being defined. The ORNL-developed Pu Disposition

Lifecycle Cost Analysis Program used for the DOE TRC Phase 1A Pu Disposition Study Report

of July, 1993 will be used as the basis for the calculations in order to integrate the results withORNL.

8.1.4 Development of Greenfield Costs versus Operating DOE Site Costs:

As a part of the Phase 1C scope of work, the ABWR Pu Disposition Study team has continued to

investigate the economic utilization of existing DOE facilities for the Pu disposition mission. As

part of this investigation, a comparison is being developed to show the relative program costs

expected if the Pu disposition complex was located at an operating DOE site as opposed to the

"Greenfield" EPRI standard Kenosha, Wisconsin location assumed for previous studies. The sites

being evaluated include Hanford, INEL, LLNL, LANL, NTS, ORR, Pantex and SRS.

Although costs developed for the upgrade and use of existing facilities are preliminary rough

order-of-magnitude costs, the comparison is expected to show that considerable savings in initial

program capital costs are possible by utilizing the existing facilities and infrastructure at an

existing DOE site.

8.1.5 Transportation Costs:

Transportation costs for plutonium feed material, new and spent fuel assemblies, and plutonium

waste materials are being developed.

8.1-3

Aotivltlel I Your

DEVELOPMENT

SAFETY/ENVIRON.

EIS/vne,'K*ml -oj_"SARConstr'uctka_

SAFETY VERIF.

FUELUCENCING

DESI(3N

CONSTRUCTION

PRE-OP TEST

OP. READINESS

STAFrFUP

OPERATION

OO __ DESIGN

,_ _RMCTION.i_ PRE-OP TEST

OP. READINESS

STAFFrUP

OPERATION

MILESTONES

Fix Env. Start- Reac- Fuel Fab Reactor

Tit. FIDDup of -for Plant Oper-- Fuel - Fuel Oper* Ilion- Fab - Load

FIG. 8-1 ABWR Pu Disposition ProjectSchedule- One Reactor-60 Years

DEVELOPMENT

ENVIRONMENTAL--

_o¢,rcons¢.tme_Jm'ts op_SAFETY.

I_tleI Tdle n Titte IIIREACTOR DESIGN

REACTOR NO. 1CONSTRUCTION

PRE-OP TEST

ORR/STARTUP

OPEP,ATK_N

REACTOR NO. 2

CONSTRUCTION

PRE-OP TEST

ORRPSTARTUP

OPERATION

OO

_, FUEL FAB PLANT i.ii.IIiiDESIGNCONSTRUCTION

PRE-OP TEST

ORP,/STARTUP

OPERATION

MILESTONESRx Env. Start- First Fuel Reactor

Tit. I RODup - Reac- Ptl--Crop. Fuel -tot - ComplMe

- Fab - Fuel

- Pit. -Load

Figure 8-2 ABWR Pu Disposition Overall ProjectSchedule- Two Reactor Case- 60 YEARS

' ,, Ill III

DEVELOPMENT i •,

SAFETY/ENVIRONM. J

EIS t

SAIl P4_r.Con=sl."Amendm'ts- ODer.,Design Constrt_cilon

SAFETYVERIF. ' '

,,. == liiiFUELUCENClM3

- REACTOR DESIGN -'qt_ 1 l_t_ It- r- _le i. -ii I

-pEACTOR.O._I " ( (ii i i-- OC_S_qUCTICN -- I Site Prep. Construct;onI I r

PRE-OP/ORWSTARTUP rLr-I-"l

CPERATI_ Z

REACTOR NO_2

(X_S.I_IUC_0N ! B

PRE-OP/ORR/STAFITUP i , , ,J

REACTORNO. 3

-- O(_IS_ [ ,!o_ PRE-OP/ORR/STARTUP ! _ r't

!OPERA'nON !

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PRE-OPTEST I_

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Rx Env. Start- Reac- Reactor _Tit. I _ "up of tor C_J_-a_,,_Corn- Fuel Fuel ComplmDplots Pal) load

I Plant II I

' FIG. 8-3 ABWR Pu Disposition Project Schedule-m

•- Three Reactor Case-25 Yearsi i l i i s a a I i l i J * i * I i t I I | | | |

A©tivltloe I Year

DEVELOPMENT

SAFETY/ENVIRONM.

EISSAR

REACTOR DESIGNREACTOR NO.1

siloPre_ConsmJc,on

ST_X:XDERATIGN

REACTOR NO. 2

STARTUPK:X_ERATIC_

REACTOR NO. 3CCNS_tJC11CN

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FIEACTOF_NO. 4

CChBmUCTICNSTARTUP_OPERATION

REACTOR NO. 5

oo STARTUP£)PE3_TR_

-- REACTOR NO. 6"..3CCNSmL_11_

STARTUPK)PERATIC_I

FUEL FAB PLANT

DESIGN

CCNSIW_'IlONST_

OPERAllCW

MILESTONESRee_or

FIG. 8-4 ABWR Pu Disposition Project Schedule-Six Reactor Case-2S Years

i

_ II lIt' I ill I i I _.

_ ii__ ' __,- ...........I-

,.--III

_" -- 0

ii __ i

.,..,,

innl n nnnll n L,.

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" .I_ ;_- -_ ........

" _ '" . _I,-, F,, 0

o ooo_ ,..._',--_ ....._, ,, ir ..............

_ _._ ,...........,_ y_o"i

8.1-8

APPENDIX A: COMPLIANCE OF MOX FUELED GE9 ASSEMBLY WITH

AMENDMENT 22 OF NEDE-24011-P.A (GESTAR II)

Introduction:

This report presents generic information relative to the GE9 fuel design employing MOX

fuel with less than 10% core average Pu enrichment and analyses of GE ABWR deploying this

fuel. The report consists of a description of the fuel licensing acceptance criteria as specified by

Amendment 22 of GESTAR II (NEDE-24011-P-A, General Electric Standard Application for

Reactor Fuel) and the basis for generic compliance of the GE9 fuel design employing MOX fuel

with those criteria. It is realized that these criteria were developed for standard urania fuel and the

criteria have to be evaluated for applicability to MOX fuel. The effect of Pu addition to the urania

fuel on the physio-chemical properties of the fuel have been descried in Section 2.1 of the main

body of this report. Urania fuel designs which meet the criteria of Amendment of 22 are approvedfor use in BWRs.

In some cases, satisfaction of these licensing acceptance criteria requires cycle-unique

analyses which must be performed after the core loading pattern for that cycle has been specified.

For those cases, the generic information contained in this appendix will be supplemented by plant

cycle-unique information and analytical results. This cycle-unique information will be documented

in a separate plan cycle-unique reload licensing report for each reload.

Copies of NRC safety evaluation report and proposed licensing criteria for fuel designs and

critical power correlations (Amendment 22 of GESTAR-II) as submitted to the NRC are available

and will not be reproduced here.

Evaluation;

A set of fuel licensing acceptance criteria has been established for new fuel designs,

developing the critical power correlation and Safety Limit MCPR for these designs, and

determining the applicability of previous generic analyses to these new fuel designs. Typically, the

fuel design compliance with the fuel licensing acceptance criteria constitutes USNRC acceptance

and approval of the fuel design without specific USNRC review. In assessing this procedure for

MOX fueled designs, it is necessary to examine each criterion and evaluate the manner in which the

fl-I

use of MOX fuel as opposed to the standard Urania fuel (with or without burnable poison) changes

the criterion,either qualitatively or quantitatively.

CRITERIA

1.1.1 General Criteria

I.I.I.A NRC Approved Models

"A NRC Approved analytical models and procedures will be applied."

Table 1, taken from NEDE-31917P shows a list of NRC approved methodologies. These models

and application procedures have been approved for licensing any new GE9 fuel design. Since the

only change pertains to the use of MOX fuel (that is, urania fuel with higher Pu content), the

general methodologies employed are not affected. In particular, all phenomena external to the fuel

rods, such as the mechanical performance of the assembly and the thermal-hydraulic response of

the assembly require no changes to the methodology. Changes might be needed in the future for

specific codes and these have been identified:

Nuclear methods, TGBLA and 3-D simulator may require additional verification for use of

MOX fuel with benchmark data to be generated from the lead testing program described in Section

2.1 of the main body of this report. Since Pu is a normal component of the material mix in the

nuclear analysis of the urania fuel where Pu is bred from U238, no new cross-sections are

required. Validation of the nuclear codes for MOX fuel will however be provided through the use

of benchmark data generated from lead MOX testing and cross-correlation with Monte-Carlo

Analysis.

Fuel rod thermo-mechanical codes such as GESTR-MECHANICAL, GECAP,

SAFR/GESTR-LOCA, may also need additional NRC approval, for example in respect to such

factors as fission gas release and pellet-cladding mechanical interaction. Validation will be

provided by the use of benchmark data generated from lead MOX testing and the predictions of the

applicable codes for the test conditions.

None of the other codes are affected.

I.I.I.B Lead Use Assemblies

"New Design features will be included in lead test assemblies."

Table I NRC Approved Methodologies

I II

References(See Attachment B)

Nuclesr

OEBLA (GENESIS) 10, 11, 12, 58

BWR Simulator(GENESIS) 12, 14, 16, 58

GEMINI Physics 67, 71, 15

Void/Doppler (GENESIS) 58, 73, 74

ExtendedBurr=._p 17, 18, 19,20, 65

Thermal Hydraulic

ISCOR (methods as described in GESTAR) 58, 15

GEXL/GETAB 41, 43, 44, 45, 46, 52, 58, 71

GEXL-Plus 50, 51, 53, 54, 55, 56, 57, 71

Safety IJmlt MCYR

OLCPOW 41, 42, 43, 47, 48, 49, 52, 71, 46ii

Transient Analyses

ODYN 1, 2, 3, 8, 9

GEMINI/ODYN 4, 5, 6, 7

SCAT/'rASC 58, 89

CRNC 1,% 3

REDY (cold water injectiononly) 81, 82, 83

GEBLA (GENESIS) 10, 11, 12, 58

BWR Simulator(GENESIS) 12, 14, 16, 58

GEMINI Physics 67, 71, 15

GEXL-Plus 50, 51, 71

GEXL/GETAB 41, 52, 58, 71

Smbmty

FABLE 21, 22, 58

fl-3

Table I (Continued)

LOCA_CS Performance

CORECOOL 71, 84, 85

, SAFER/OESTR.LOCA 65,67,76,77,78,S4,S5SAFE/REFLOOD 31, 32, 67

CHASTE 31, 32, 33, 37, 67

SCAT 31, 32

LAMB 31, 32

GEGAP 86

Clad Balooning 31, 32, 34, 35, 36, 37, 65

ATWS

REDY 87,SSTASC 87, 88

STEMP ST,SSVoid/Doppler 58

RodDrop Accident

"GE RDA calculationmethodology" 24, 70, 91, 92, 93

BPWS 23, 80, 94

FuelStorage andHandling

Spent fuel storage methods 23, 64

Fuel handing accident 30

Thermal.Mechanical

GESTR-MECHANICAL 24, 65

Seismic and LOCA loads 79

Extended Burnup 17

Creep-C_llapse

Bowing 28

R-4

Compliance with this criterion has been fully met. No new design features are employed in the

proposed GE9 fuel design employing MOX fuel.

I.I.I.C Post-Irradiation Fuel Examination:

'The generic post-irradiation fuel examination program approved by the NRC will be maintained."

NRC-approved fuel examination program will be used in implementing the rod tests described in

Section 2.1 of this report on lead testing program. This will include lead detection tests, such as

sipping, visual inspections, nondestructive testing of select_ rods by ultrasonic and eddy current

techniques and dimensional measurements. In addition, axial gamma scanning and destructive

examinations for fuel metallography will also be carried out on selected rods.

1.1.1.D New Fuel-Related Licensing Issues

"New fuel-related licensing issues identified by the NRC will be evaluated to determine if the

current criteria properly address the concern; if necessary, new criteria will be proposed to the

NRC for approval."

The major new fuel-related licensing issue is the use of MOX fuel. A qualitative assessment of the

effect of up to 10% Pu addition to the standard urania fuel on fuel properties and performance has

been carded out in Section 2.1 of this report. Based on this initial assessment, it is concluded that

no new fuel-related issues hz 'e been raised by the use of MOX fuel.

While the properties of the fuel and the thermo-mechanical response are affected by the addition of

Pu, it is concluded that these effects are minor for Pu enrichments to 10%. For the levels of core-

averaged enrichment levels of Pu proposed in this study, which are nearer to 5%, no changes to

the current methodology or criteria are envisioned. The required modifications to the fuel

properties are readily incorporated based on available data, and are imbedded in the current models.

Satisfactory performance of MOX fuel in this range of Pu enrichment has already been verified. A

more detailed assessment will be carried out during the following phases of the study to confirm

this finding. The results of early MOX rod tests proposed in the lead testing program will also be

used to confirm these findings.

I.I.I.E NRC Separate Review

[I-5

"If any of the criteria in Subsection 1.1 (of Amendment 22) are not met for a new fuel design, that

aspect will be submitted for review by the NRC separately."

All the criteria of Subsection 1.1 are met, however, NRC will have to independently evaluate the

findings given in Section 1.1.1.D above.

1.1.2 Thermal.Mechanical

The subsections which apply to fuel rod thermal-mechanical design are:

i. "The fuel rod stresses, strains, and fatigue lifetime usage shall not exceed the material ultimate

stress or strain and the material fatigue capability."

vi. "Loss of fuel rod mechanical integrity will not occur due to excessive cladding pressure

loading."

ix. "Loss of fuel rod mechanical integrity will not occur due to fuel-melting."

x. "Loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical

interaction."

All these limits, with the exception of those related to fuel-melting, will be verified using (a)

analytical models and the extensive data base already available for MOX fuel properties and

performance, and, (b) by the results from the lead MOX rods tests recommended in Section 2.1 of

the report. The principal reason for the tests, even though a significant MOX fuel data base is

already available, is to fabricate fuel to given QA requirements to produce the specific fuel

microstructural features needed to ensure satisfactory thermo-mechanical performance of the fuel

rod. Fuel-melting related issues will be addresses through a combination of already available data

base as well as new thermal arrest studies, should they be warranted.

1.1.2.B (ii)Fretting, 1.1.2.B (iii) Metal Thinning

"Mechanical testing will be performed to ensure that loss of fuel rod and assembly component

mechanical integrity will not occur due to fretting wear when operating in an environment free of

foreign material."

11-6

"The fuel rod and assembly components evaluations include consideration of metal thinning and

any associated temperature increase due to oxidation and buildup of corrosion products to the

extent that these effects influence the material properties and structural strength of components."

These criteria will be met by the GE9 design and the performance is unaffected by the use of MOXfuel.

1.1.2 B (iv) Fuel Rod Internal Hydrogen Content

"The fuel rod internal hydrogen content is controlled during manufacture of the fuel rod consistent

with ASTM standards C776-83 and C934-85 to assure that loss of fuel rod integrity will not ,.:cur

dueto internal hydrating."

Compliance of this criterion, which controls the maximum amount of hydrogen allowed to be

present in the manufactured fuel rod, will be demonstrated as part of the lead testing program.

1.1.2. B (v) Fuel Rod/Channel Bow, 1.1.2. B (vii) Control Rod Insertion, 1.1.2.

B (viii) Cladding Collapse

"The fuel rod is evaluated to ensure that fuel rod or channel bowing does not result in loss of fuel

rod mechanical integrity due to boiling transition."

"The fuel assembly (including channel box), control rod and control rod drive are evaluated to

assure control rods can be inserted when required."

"Loss of fuel rod mechanical integrity will not occur due to cladding collapse into a fuel rod

column axial gap."

These criteria are unaffected by MOX usage and will be met by the GE9 design.

1.1.3 Nuclear

1.1.3 A Doppler Reactivity Coefficient

" A negative Doppler reactivity coefficient shall be maintained for any operating conditions."

This criterion has been demonstrated using approved methodology for all operating conditions.

1.1.3 B Moderator Void Coefficient

1t-7

"A negative core moderator void reactivity coefficient resulting from boiling in the active flow

channels shall be negative for any operating conditions."

This criterion has been demonstrated for all the core design options, using NRC approved

methodology.

1.1.3. C Moderator Temperature Coefficient

"A negative moderator temperature coefficient shall be maintained for temperatures equal to greater

i than hot standby using NRC approved methodology."

This criterion has been demonstrated using NRC approved methodology.

1.1.3. D Prompt Reactivity Feedback

"For a super prompt critical reactivity accident (e.g., control rod drop accident) originating from

any operating condition, the net prompt reactivity feedback due to prompt heating of the moderator

and fuel shall be negative."

The ABWR design features preclude accidents such as control rod drop. Reactivity feedback will

be dominated by the prompt Doppler feedback and the margins for a postulated reactivity insertion

is expected to be as good, if not better than urania fueled cores. Analyses will be carried out to

quantify this margin during the detailed phase of this study.

1.1.3.E Power Coefficient

" A negative power coefficient, as determined by calculating the reactivity change due to an

incremental power change from a steady state base power level, shall be maintained for all

operating power levels above hot standby."

The power coefficient is derived based on three component feedbacks, Doppler, Modertor Void

coefficient and Moderator Temperature coefficient. Since all three are negative, no further analysis

is required.

1.1.3.F Cold Shutdown Margin

"The plant shall be calculated to meet the cold shutdown margin requirement for each plant cycle

specific analysis."

The required cold shutdown margins for the ABWR plant has been demonstrated using NRC

approved methods.

I.I.3.G Fuel Storage

" The effective multiplication factor for new fuel designs stored under normal and abnormal

conditions shall be shown to meet fuel storage limits by demonstrating that the peak uncontrolled

k-infinity calculated in a normal reactor core configurationmeetsthe limits provided in Section 3

(of GESTAR-II) for GE-designed regular or high density storage racks."

The k,,. of the proposed MOX fuel designs is very similar to the urania fueled designs (there are no

system level changes) and therefore all requirements for fuel storage will be met as easily. Detailed

analyses will be carried out in the future.

New Fuel Design Licensing Evaluation (No corresponding subsection in

Amendment 22)

"Licensing evaluations of new fuel designs will include generic analyses of the ABWR plant at

limiting points of the cycle for an equilibrium loading of the new fuel design to assure that (1)

nuclear design criteria are satisfied, and (2) safety limit MCPR values are correct. In addition,

Chapter 15 safety analyses are performed for each reload application on a cycle-specific basis for

(3) limiting anticipated operational occurrences and (4) bounding accidents. The cycle-specific

plant (5) operating limit MCPR is determined and the effect of the new fuel design on previously

evaluated accidents must be reconfirmed or reanalyzed."

Compliance with the nuclear design criteria and safety limit MCPR has been documented at limiting

points of the cycle for equilibrium loading for each of the core design options presented in the main

body of the report. Compliance with cycle-specific anticipated operational occurrences, bounding

accidents and operating limit MCPR will be performed in the more detailed phases of this study

when cycle-specific data are generated.

1.1.4 Thermal-Hydraulic

" Flow pressure drop characteristics shall be included in plant cycle specific analyses for the

calculation of the Operating Limit Minimum Critical Power Ratio."

The flow pressure drop characteristics will not change as a result of the use of MOX fuel and these

will be included in cycle-specific analyses during the detailed phases of this study when cycle-

specific data are generated.

1.1.5 Safety Limit MCPR

I.I.5.A Confirmation of Applicability

"Safety limit MCPR shall be recalculated following steps in 1.1.5.B (of Amendment 22) or

confirmed when a new fuel design or new critical power correlation is introduced."

The safety limit MCPR is influenced by the critical power correlation and by bundle design

parameters which affect the bundle R-factor distribution and the core radial power distribution.

These parameters include the spacer design, assembly dimensional geometry, enrichment level and

distribution, and fuel discharge exposure. The spacer design and assembly dimensional geometry

remain the same as for conventional urania designs. For the GE9 fuel design, the recalculation of

Safety Limit MCPR will follow the steps specified in Subsection 1.1.5.B of Amendment 22 asdefined below.

1.1.5.B Safety Limit MCPR Calculation

"Safety Limit MCPR calculation will be performed under the following conditions."

i. Analysis shall be performed for a large high power density plant.

ii. Analysis shall be performed for a bounding equilibrium core.

iii. Core radial power distributions shall be selected to maximize the numberof bundles at or near thermal limits.

iv. Local power distribution shall be selected such that the largest anticipatednumber of rods will be near boiling transition.

v. Ninety-nine and nine-tenths percent (99.9%) of the rods in the core mustbe expected to avoid boiling transition.

vi. Uncertainties used in the analysis shall be the same as documented in GESTAR I1,Section 4, except for the uncertainty associated with a new critical powercorrelation. The critical power correlation uncertainty used in the SafetyLimit MCPR determination, shall be that uncertainty associated with theoperating regions that can be obtained during normal operation or duringanticipated operational occurrences."

A-18

Safety limit MCPR is not expected to change as a result of the use of MOX fuel. However, withany new fuel design, it is customary to recalculate the safety limit MCPR and this will be carriedout for the specific MOX fuel designs (to be chosen) for disposition.

1.1.6 Operating Limit MCPR Licensing Evaluation

I.I.6.A Cycle.specific Analyses

"Plant operating limit is established by considering the limiting anticipated operational occurrences

for each operating cycle."

Cycle-specific analyses will be conducted during the detailed phases of the study when such cycle-

specific information will be considered.

1.1.6. B Generic Analyses

"For each new fuel design, the applicability of generic MCPR analyses described in Section 4 (of

GESTAR II) or in the country specific supplement to this base document shall be confirmed for

each operating cycle or plant-specific analysis will be performed."

Operating cycle/Plant-specific analyses will be performed during the detailed phase of this study.

The applicability of the MCPR analyses will be demonstrated.

1.1.7 Critical Power Correlation

I.I.7.A New Fuel Design Features

"The currently approved critical power correlation will be confirmed or a new correlation will be

established when there is a changed in wetted parameters of the flow geometry; this specifically

includes fuel and water rod diameter, channel sizing and spacer design."

No new fuel design features involving the wetted perimeter or other hydraulic parameters arise as a

result of using MOX fuel.

1.1.7.B New correlation data

"A new correlation may be established if significant new data exists for a fuel design."

No new c,orrelation arises as a result of using MOX fuel with the GE9 bundle design.

1.1.7.C Critical Power Correlation Calculation

Not applicable as no new correlation is involved.

A-II

1.1.8 Stability Licensing Acceptance Criteria

"The new fuel design must meet either of the criteria described below:"

I.I.8.A Comparison with previously e Jproved designs

"The stability behavior, as indicated by core and limin'ng channel decay ratios, must be equal to or

better than a previously approved GE fuel design."

Stability analysis has been conducted for the proposed designs and the allowable flow-power

operating region during startup set to satisfy this criterion.

1.1.9 Overpressure Protection Analysis

"Adherence of the ASME overpressure protection criteria shall be demonstrated on plant cycle

specific analysis."

Bounding analysis confirming that the overpressure protection criteria has been met is reported in

the main body of the report for the equilibrium cycle conditions. Cycle-specific analyses will be

carried out when such information is considered during the detailed phase of this study.

1.1.10 Loss.of-Coolant Accident Analysis

1.1. IO.A Emergency Core Cooling System Criteria

"The criteria in IOCFR 50.46 shall be met on plant-specific or bounding analyses."

The emergency core cooling system (ECCS) criteria in 10CFR50.46 are met by the expsoure-

dependent maximum average planar linear heat generation rate limit in bounding analyses.

Compliance has been demonstrated for the proposed fuel designs. GE will continue to

demonstrate compliance with these ECCS crtieria for specific fuel designs used in the disposition

project.

I.I.10.B Plant MAPLHGR

"Plant MAPLHGR adjustment factors must be confirmed when a new fuel design is introduced."

A-12

PlantMAPLHGR issometimesadjustedfora specificoperationalconfigurationorregion.GE

will confirm the revised MAPLHGR limit before each cycle operation.

1.11 Rod Drop Accident Analysis Licensing Evaluation

Rod drop accidents are precluded by ABWR design features.

I.II.A Cycle Specific Analysis

"Plant cycle specific analysis result shall not exceed the licensing limit described in GESTAR H."

(Plant cycle specific analysis will be conducted to show that the Ganged Withdrawal Sequence for

the specific reload is in conformance with all accepted limits. However, rod drop accidents are

precluded by ABWR design features.)

1.1 I.B Bounding BPWS Analysis

"Applicability of the bounding BPWS Analysis must be confirmed."

(See comment under 1.11.A)

1.1.12 Refueling Accident

"The consequences of a refuel accident as presented in the country-specific supplement or the plant

FSAR shall be confirmed as bounding or a new analysis shall be performed (using the methods

and assumptions described in the country supplement) and documented when a new fuel design isintroduced."

The activity released to environment during a refueling accident is principally related to the

exposure and not the Pu content. Therefore, the consequences will be similar to a urania fueled

core and compliance with applicable limits will be demonstrated.

1.1.13 Anticipated Transient without Scram

"The fuel must meet either of the acceptance criteria described below:"

A-13

1.1.13.A Void Reactivity Coefficient Range

"A negative core moderator void reactivity coefficient, consistent with the analyzed range of void

coefficients shall be maintained for any operating conditions above the stratup critical condigtion."

The GE9 design has been generically analyzed and found to satisfy this criterion if the void

coefficient falls within the range of -2.5 to -11.6 cents/% voids. The maxium void coefficient for

the reference fuel design case was -13 cents/% voids, slightly more negative than the range over

which generic analyses had earlier been conducted.

1.1.13. B Plant Evaluation

"If the preceding criterion is not satisfied, the limiting events will be evaluated to demonstrate that

the plant response is within the ATWS criteria."

Background analyses for the ABWR have shown that the ATWS presuurization transients are not

the most limiting. Specific overpressure bounding transient analyses have been conducted

showing compliance will applicable peak vessel pressure limits. Such a bounding transient

involves the sudden closure of Main Steam Isolation Valve (MSIV), the normal MSIV position

scram is assumed to fail - the reactor scrams on high neutron flux. The details of this analysis are

contained in the main body of thi_ report.

R-14

APPENDIX B: REPOSITORY CONSIDERATIONS:

10 CFR PART 961, Article Vl. Criteria for Disposal

__poe_0.

A. Qmmsd J_qwtmUt. _ d. I_el may trove "5PatJedreef* mxt/armwnd "5__ Fuel" ealUlmt_v(s) _zospt u _ i_ed In t/sis

emu'_ DOu ,_ uomt hereunder ULJy a. _ _ end8uh:hue_SeeU_--su_ BIO' ud/_.i.K,W.._eh meow t&eOermx_l0_mtJl_n, for _ _ amd Oemuy_,lpael/lma_wuW u mt for_ lnaM_adlx B, anneud 1. _ Now_e_ P_stcal fHwum.liereto imd msl i_pert tus_of.

Cn) _ 8tmn ueurdely elmOrlINT and/or HLW _ to delivery In ee. i| |

L ,Pl_n_. -_ ...........

taned dmnc_e ot the 81f end/or Him Je_, _,_,_ u _ ,J_,.--I ,, J,,_o_,L_i_____ Ol8 lieaJ0ns-_ IJInctm8X0 iSSuemuten_--m¢ forthtn_ r. mmend ..... J_,L ] _,0tunto sad msde a mr_ ttmwt, Pun_sssr____Rd__ eueu,uKtn mid 81Q' and/orIRZ_Wu man u ttzey nmu:_ al i_=umleumbecome tnewn to me pm_tum_, w _ m 0_mm_Jr_-_m ._ um_j

Co)DOB'8eblilptlou lot' dbpmtn8 of Nun,

than etendud fue_ herons, tot 8m__ nen4s tn_udtns, but not ISmtt_tto, eentroli _ been__ bY the _ meiden_burnable polmmrod mmemblJm,ran.eruothertbms4mubrdfuei, uUm_teem I_l rod elemeatL Udmble _uSm, filmb dsdlned In _ _. the _ etmmoe_ ud mtmary end meoadu_ neu-lli____ trmssom_ tlmt s_ amtma_ mtl_ tim

mtULUedviae Purehmer u_bin jUt_ (04)) Intelrad _ of the _ miemtly. _mktt douo_ _eqldre _ tnendt_, rely _ Jaetud-_ _r _ _ _ _ r_queerN to tho teeeddcslfmfutllty of dis. ml u 94trtair tho spent nvrJeu'hid de/h.erul toe 6i/mml mnmmn¢qtolid_ eonlru_IPOSLn_of amebrue1m_tho ems,eull_ ,nereed IPle_z:Puml_ does,,oil,m_etUNt au_ecJ.

let eueb nr54e_ Puei--CluJ NS-L_s.tm_z E a, Ceeft_. _ ndntmumeooU_ ttme for

fuel _ live (8)Oeae_/,Ir_io_a N_ _ _ dim notmeetthis8pecUt.

es_lon 868_ be etmmifled 88 Hmwhmdardd, _ _ ldilsltfluiffos _

1. _o__ s_ usereuon. _ _o_ _ _el _ e_er _umlibleeflor_t, utUIMn8_ equlssde_rt, II_R pogey llclllUq_ IdudJIbe dMmlftedhis_. _o_ _ _ __ peru. N_ __i ]18.4. 8ueh fuel

flee. to _ dsm_ 80_t N_ _ _ be _ue and _ 8eee_! tumdlmur.

(81q_) prior to deMver_te DOE, u lol/omcaKoiwse,and l lamlllUee,

_mul_l JPaaelmmun 6N_Pthat mee_ 11.(3mup_ _ Roal_IPWlwhichhuilL1the 131_rterld_l_m __ wt beendll_aembled8_d 8Lmredw4U3LhefualIov_hIn pau'snrrsphB belew, rods In • eoeuoik_ted rammer_QI _

laOtmeet otw or unoveof th_ Oeneml _I- e. P_8_fedYueLse_ focth U_ 8ubl_rsSr_J I 8, Vlsual I_or_

threuSb 6 o! _h B bdou. gmcli __ _ bev_tadl_, _ for18lid S NU_ _el _ evidenceof shrucq_nd defo_ _ _eMH-ItJ_ou4_ _ pmmmnt to _ to timing or m_ers uldeh n_3_ rectuireB below, slmdld haindllhll.Assemblies whJei_(|l 8xe

C, ,eaffed _ me_r_ 8NF Um_ mNtt, _ 81a3_e_Undlydoformedor hive danmgedspecaflmUons ut forth tn 8ub_smemptu I et_dtr_ W _ extent tJ_st_il _QridlingII_h $ ol ps_ph B below, Jmdi _ be usqulred or Ill] for 8my _n

_ u _ _I _ _-L _ be _ed _th _tl I_I him.U515ml_F4 _t tO___ e of _ equlprnerlt_tll be ehum_fiedlUl._I_penGrnph B Ibelow. IPtx__ F-I.

B-1

is. Ptmvimss_ _ted Ammmbitee. UL _ Jut], dlatorttom, cinddtnz,/bmRblies eueupmJle_ed by _ deme_ m. o_er dmmmle to the spent,fuel,

pcJor to cli.uUkmUon hefmmdlr Jhd! k or nonfud eoatl_nenla wit.bin U'.ts 8h_pirtirdtJdtlul u _ lqLel--OuJ r-8. Put- bit W_ will rtqulre JiJectt_ hudilnt J_m.ebsewr shaU sdv_ OOB of the _mson for eedurea. (Al_,ach 84fdft0OllA|pales if needed.)the I_m' enespsudsUon of Msemblleu in go/6 ........liclen4 detail so _ DOE may plan for 8_ = .......prOl_ldm attbaequel_,hJxldlk_, IV, Amemb4y-iRlimber----

o. Relulsto_ Requlresmmta. 8hlpotn8 lot #---8peril fuel mm.mbUes _nsll be pedmaed

end nbzed In etmkn so thst, sll sppil_blereaulutory requlrememt8sre reeL. . .

1./Jlandmnf PuzL' • l_L CIm8 8-1: -I--l.--l.-.l.--l-..-b. Ctus 8-1: BWR _.8tmw iw (mwew_)1. Nmukndm_ JL_j_r: t._ em (.wcJm_ '!".... !--.-t-.--'t ..... |'--

s. el..8.-I: _lcul Dlme_,, 8._. __....J_ l 1 )b.am Ns-_,J_on_ com_ner_ 4._ _ row,,ms**_c.:..j.__.C:: "e. Cbt.u H6-3: 8holq, Coeled J, ImaJ_ eetm_tLmmmJyInm_ v_ _ qqm_Jd. Clam Ha-4: Hon.bW'R mWelaml meu_t ------- m e_0me ------e. Cims N8-5: Como.d_ted 3PuelRods. .......S./FeU_ FUEL"it,.CIm F-U VJmml PaJlurrt e_ I)sma_ Any lslSeo n_lUou8 or frsdulenL at_te-b. Ctus I:'-_ Rsd_os_We "16esk_e" men&m_¥ be plmhdltble by fine or ImDrls.

Ch._ P-3: ICncspmJst,od oemumt (U.8. Code. 'r'iUe L8. l_t, ton lOOI).

D, HWh.Lnel RadCa_t(_ Weufe B_ Pur_luumn

The DOE st_dl ecce_ tdah.levet rmdtotc. 8la'naCureWe wute. De_lled 8ece_ta,nce er_14t, and Title8enera/a, pecU'(cst,ions for tmch waste will be DateJsmsed by Che DOE no later thJm the dMe on_hh:h DOB sulbmlla 114llcetsse o4ppl_c_UontO the Nuclear Rel_at,0r_ C_Jm_sdo_ forUse fln_ dtspoeal IselllLy.

'Am,mmaz P

/)e£eUed/)esc_pfh_4t oy'PuyrTt4serV

lmrormsMon Mudl be jurm4ded byl_rcJ_=er for esch dieUn_ fuel type _t&hln• 81hlpplrurLo_ no(. later tlum 81xf,y (80)cht,ym prior Lo the _edule LmM_rUt_Joodate,.Purehme_Con_nu:&Number/I_te ----t--------Rnctor/lP_tcJllt, Y _sme--

I. D_wln_ tn,'_uded _ 8enertc doutec.mmmmm._Im

t. lPuel ,As_.mt_ DWGI ---2. ODl)e_& l_mer ewd llt,Unls DWGI --Domt_ lqumber: ---DOE _bl4ppinf:._tf: -----I AmembJtesDeecrlbed:

...---PWR

.----OUner

[I. Deskm M_'#_ Desc_Mions.Jh_dJteme_tL"

1. Elemen_ tylX ----(Js_, plal,e, etc.)I. _04sl ler_h' --/tin.)L Active lenl_ --(in.)4. C]tddlng mttertal -- (7_",8.8.. etc.)

,JuemMv Deecr(_tton."

I. Humber o! Etemen_ ----2. Overall dtm.-,t_ons (le_nl£h ---- (cross

Imct,ton)---- (in.)_I.O_crld] veill_r.. --..-

B-2

BWR SPENT FUEL DISCHARGE STATISTICS

"T"_t.._ .S:2. Histoncallydischargedas_mblicsforthe GE BWR/4-6 Aue.mbly_ brown downby AssemblyTypeandDischargeYear (reproducedfrom the LWR QuantitiesDatabase).

i i i

LWR QUANTITIT_ DATABASE

Historical Data through December 31, 1990DataBrokenDown By:.Assembly _, Dis,-J_,geYearDischargedAssembliesby Aue.mbly Class:OE BWI_4-6

AVEI_GE TOTAL AVERAGEDISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIAL

YEAR ASSEMBLIES ASSEMBLIES" (Mwd/MTIHM) (M'II}IM) ENRICH.,|, li ' i ,

Assembly Type: GE BWR/4.6 7 X 7 GE-21973 50 50 3741 9._ 2.5001974 328 328 9199 63.3 2.5001976 285 5 g718 55.8 1.0981977 304 1 9848 59.5 1.1071978 137 1 9620 26.8 1.0981979 38 10747 7.4 1.097

Assembly Type Totals 1142 585 9373 222.4 1.559

' Assembly Type: GE BWR/4-6 7 X 7 GE-3a1975 2 2 4653 0.4 2.1201976 104 19 7947 19.5 1.3891977 397 21 14032 74.4 1.9931978 575 24 20874 107.8 2.4101979 462 10 18787 86.6 23121980 1233 _ 22 23312 231.2 2.4341981 490 8 24304 91.9 2.450

1982 337 4 23143 63.1 2.2941983 12 22942 2.2 2,334

1984 64 24425 12.0 23341985 76 21396 14.2 2334

Assembly Type Totals 3752 110 21057 703.3 2325

AssemblyType: GE BWR/4.6 7 X 7 GE.3b1976 191 3 10570 36.4 2.0561977 1 19646 0.2 2.5071978 312 3 20538 59.1 2.1501979 435 14 254545 82.6 2.4141980 112 26645 21.3 2.5071981 g7 25146 18.4 2.5041982 36 26674 6.8 2306

AssemblyTypeTotals 1184 20 21943 224.9

AssemblyType: GE BWR./4.68 X 8 ANF1989 220 26708 38.8 2.7411990 512 29675 90.4 2.863

At_cmbly TypeTotals 732 28784 129.1 Z826

• As reported _ theutiliucs

B-3

T,,blea..ti iiii i __ ' i i

LWR OUANTITgES DATABASE

Historical Data throup December 31, 1990Data Broken Down By:. Assembly "I_e, Ditclutrse YearDischarged Assemblies by Assembly _ GE BWR/4.6

AVERAGE TOTAL AVERAGEDISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIAL

YEAR ASSEMBLIES ASSEMBLIES" (MWd/MITHM) (MTIHM) ENRICH.l I ' l Jl i i J i i ii

Assembly Type: GE BWR/4.,68 X 8 OE-4,a1977 112 189'24 20.6 2.19019'78 9'2 8 18801 16.9 2.2381979 158 56 20073 _.0 2.3921980 267 73 _1 49.1 2.6701981 392 19 24825 72.2 2.6501982 271 8 26187 49.9 2.7351983 300 2 27497 55.3 2.7361984 333 10 28652 61.2 2.7631985 19 24842 3.5 2.730

Assembly Type Totals 1944 176 24993 357.6 2.631

Assembly Type: GE BWR/4.6 8 X 8 GE-4b1978 3 2 13892 0.6 2.1921979 137 4 17153 25.6 2.1771980 621 19261 116.0 Z1641981 479 22602 89.5 2.4221982 262 2 22328 48.9 2.1941983 91 29507 17.0 2.6851984 146 l 25220 27.3 2.7391985 48 16416 9.0 2.114

Assembly Type Tomb 1787 9 21368 333.7 2.311

Assembly Type: GE BWR/4.6 8 X 8 GE-51980 78 2 2848 14.3 0.7621981 33 12 20093 6.0 2.6551982 220 44 22088 40.0 2.4871983 950 18 27626 173.9 2.7061984 772 27 27108 141.3 2.6781985 630 1 19046 115.5 2.0351986 48 27421 8.8 2.2631987 264 12636 48.3 1.5461988 644 23423 117.8 2.2561989 343 27862 62.9 2.5481990 216 23405 39.5 2.123

Type Totals 4198 104 23645 768.3 2.362

reported _ the utilities

B-4

Table _1. (continued)II I ii i iBij I III I II [] II

LWR QUANTITIES DATABASEH_toncal Data thmu_ December 31, 1990

Data Broken Down By:. Auernbly 1313e, Dicflarge YearDischarged Asu:mbliesby Assembly_ GE BWR/4.6

AVERAGE TOTAL AVERAGE

DISCHARGE NUMBER OF DEFECTIVE BURNUP WEIGHT INITIALYEAR ASSEMBLIES ASSEMBLIES* (MWd/MTIHM) (MTIHM) ENRICH.

....... i i ! I I I II

Assembly Type: GE BWR/4.6 8 X 8 GE-61981 26 2 23200 4.8 2.8831982 1 9565 0.2 Z8611983 198 4 7,6833 36.2 2.8061984 556 39 18516 101.7 2.8181985 1437 96 23946 262.7 2.6261986 1231 14792 225.9 1.7681987 2583 18"/08 473.0 2.362

1988 1263 _ 3 24977 231.3 2.462

1989 1329 ._j/ 20432 243.92.O25

1990 929 20432 170.9 2.037

Assembly Type Totals 9553 144 20394 1750.6 2.297

sembly Type: GE BWR/4.6 8 X 8 GE-71984 1 1 5508 0.2 2.6571986 8 16829 1.5 2.6591987 225 8910 41.7 1.2871988 322 1 15369 59.4 1.9361989 1417 17 18086 261.7 2.2271990 1329 5 24649 245.2 2.474

AssemblyType Totals 3302 24 19826 609.6 2.235

AssemblyType: GE BWR/4-6 9 X 9 ANF Prepress.1989 1 24000 0.2 3.319

,Assembly Type Totals 1 2,4000 0.2 3.319

TOTALS 27595. 972 21094 5099.6 Z309

reported by. the utilities

B-5

Table :_. Projectedquantitiesof spentfuel from GE BWRI4-6 A,u=mblyClass,brokenclownbydischargeyear andbumup bin (Reproduced from the LWR QuantitiesDatabase)

III IN NI ummlll I I I I IN I ii

LWR QUANTITIES DATABASE

Projected Data: No New Orders Casewith ExtendedBumupDam BrokenDown By:. Di_luu'ge Year and Bumup BinProjectedAssembliesfor AssemblyCam: GE BWR/4.6

NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP --Oi_ BURNUP _r.IGHT INITIAL

YEAR BXN ASSEMEiLmS (MW_) (M'rmM) ENRXCHMENTm INI u I I n I n I n I I Ul I

1991 0- 5000 4 3000 0.7 0.7111991 , 15001-200(X) 132 17310 23.5 1.6501991 20001-25000 572 23007 102.1 Z3001991 25001-30000 I 100 28359 200.8 2.6951991 30001-35000 879 33251 157.2 3.0821991 35001-40000 4 36000 0.7 2.990

1992 O-5000 52 4077 9.2 0.9401992 10001.15000 96 13327 17.0 1.4161992 15001-20000 224 18438 39.5 1,7281992 20001-25000 $84 22416 103.7 2.2081992 25001-30000 610 28392 111,2 2.5991992 30001-35000 1421 32791 252.5 2.9921992 35001-4{X_ 113 36381 20.1 3.201

1993 20001-25000 176 24122 31.8 1.8681993 25001-30000 788 28792 143.4 2.6721993 30001.35000 1597 32959 286.7 3.0601993 35001-40000 96 36000 16.6 3.310

1994 15001-20000 180 17600 31.9 1.9501994 25001-30(X)0 768 28610 138.8 2.8061994 30001-35000 1449 32816 258.5 3.0641994 35001-40000 575 37059 102.5 3.167

1995 25001-30000 264 29143 48.0 2.7291995 34X)01-35000 200I 33192 360.5 3.050

1995 35001-4(XX)0 375 36128 67.2 3_,324

1996 25001-30000 656 29140 118,9 2.8341996 30001-35000 1114 32882 198.6 3.0441996 35001-40000 1268 36340 224.8 3.2131996 40001-45000 4 42000 0.7 3.140

1997 15001-20000 36 16000 6.4 3.2001997 25001.300(X) 79 28206 14.0 2.8451997 3(X)01-35000 1717 33525 309,7 3.0911997 35001-40(X)0 1322 37011 233.9 3.267

1998 25001.34XXX) 208 27629 37,7 3.0331998 30001-35000 676 33538 121.7 3.0411998 35001-4(XXX) 1666 36476 301.7 33001998 40001.45000 144 41004 25.6 3.620

B-6

Tmc _ (mntin_)II gl ii I ii I ii I II I I I I III II III II I

LWR QUANTrrIF_ DATABASE

Projected Data: No New Orders Case wills Extended BumupData Broken Down By:. Disct_'ge Yea' and Bumup BinProjected As_mbliesfor Assembly_ GE BWR/4.6

NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INFIIAL

YEAR BIN ASSEMBLIES (MWd/MTIHM) (MTIHM) ENRICHMENTII I Ill I I I II I I I I

1999 15001-213000 179 19155 32.4 2.5101999 25001-34X}00 55 28562 9.7 3.0111999 30001-35000 1312 33702 235.1 3.1131999 35001-40000 1990 36638 354.4 3.209

2000 25001.30000 219 28024 39.4 3.0622000 30001-35000 478 33399 86.5 3.2282000 35001-40000 1835 36977 328.8 3.272

2001 25001.30(D0 198 29950 35.9 2.6112001 30001-35000 385 33067 70.2 3.1172/301 35001-40000 1616 37293 289.7 3.2822D01 40001-45000 136 42066 24.3 3.687

2002 25001.300(X) 147 27931 26.7 3.0832002 30001-35000 238 32,440 43.2 3.2642002 35001-40000 2532 37691 455.3 3.3462002 40001-45000 557 40607 99.0 3.361

2003 25001.30000 92 28539 16.5 3.1012003 30001.35000 91 33589 16.3 3.3722003 35001-40(_ 1667 38031 298.1 3.38521303 40001-45000 514 41087 91.6 3.3562003 45001-50000 129 45039 22.9 3.874

2004 25001-30000 149 28315 26.9 3.0772004 30001-35000 264 33177 48.2 3.1852004 35001-40000 1721 38103 309.4 3.3022004 40001-45000 223 40789 40.0 3.575

2005 25001-30000 73 28720 13.1 3.0992005 30001-35000 310 33272 56.2 3.23821305 35001_ 1993 37709 357.5 3.3282005 40001-45000 1221 41297 215.2 3.509

2006 25001-3(X_ 145 28271 26.3 3.1042006 30001-35000 110 34122 20.0 3.54421306 3500 I-4(X)(K) 1439 38228 262.3 3.3332006 40001-45000 283 42626 51.5 3.678

2007 25001-300(0) 73 28938 13.2 3.1142007 30001-35000 185 32627 33.3 3.2642007 3500 I.-40000 1618 37384 289.3 3.2892007 40001-45000 1058 40925 186.5 3.510

B-7

Table'S2,. (cominu_)

L_t QU_ DATABASE

ProjectedDam: No New OrdersCasewith E,s'tend_BumupData Broken Down By: DisclmrlteYear zud Bumup BlaProjected A.uanblies for Assembly Class: GE BWR/4.6

NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INITIAL

YEAR BIN ASSEMBLIES (MWd/MI]HM) (MTIHM) ENRICHMENT

2008 25001.30000 147 28547 26.7 3.0922008 30001.35000 194 33642 33.3 3.2852008 35001.40000 2097 37967 378.9 3.3532908 4000145000 913 40985 1617 3,5732008 45001-50000 128 45,,165 22.7 3.901

2009 23001-30000 73 _ 13.1 3.1222009 30001-35000 II I _ 20.0 3.1932009 35001..40000 1303 311147 233.4 3.2972009 40001-45000 29 41310 5.5 3.359

2010 23001.30000 141 29221 25.5 3.1642010 30001-35000 202 32951 36.4 3.2812010 35001.40000 2020 38101 365.5 3.3522010 40001-45000 1427 41145 232.7 3.5122010 45001.50000 127 45800 22.6 3.922

2011 23001-30000 142 29416 23.7 3.2052011 30001-35000 44 34214 &0 3.20320 11 35001-40000 1081 38256 195.2 3.39920 11 40001-45000 315 42117 56.7 3.404

2012 15001-20000 132 16149 23.5 3.1872fi12 30001-35000 272 30958 49.0 3.2802012 3500144X_ 497 38619 89.7 3.4492012 40001-45000 1810 41647 324,,5 3.561

2013 10001-15000 228 13349 41.4 19712013 20001-25000 80 22262 14.5 3.0482013 23001-30000 148 26255 26.9 3.0822013 30001-35000 232 31593 411 3.1702013 35001-40000 759 37903 138.0 3.4032013 44)001-45000 749 42575 134.4 3.4942013 45001-50000 129 45128 22.8 3.596

2014 5001-104_ 28 9457 5.1 3.2172014 10001-15000 716 13694 128.7 3.2852014 15001-204_ 472 16632 84,.0 3.3632014 20001-25000 144 23879 23.6 3.3192014 25001-304_ 840 28250 150.8 3.4452014 30001-35080 324 32708 57.8 3.4272014 35001-404X_ 881 38378 157.3 3.4862014 40001-450430 985 42159 176.9 3.6262014 45001-50000 2/30 47572 35.2 3.903

B-8

Tree_2.. (cm_ucd)i i,i ii i ..... i i, i i iii i i i,

LWR OUANTITIES DATABASE

Projected Data: No New Orders Care with Extended BumupData Broken Down By:. DltclutrSeYear and Bumup BinProjected_blk_ forAssemblyClass:GE BWR/4.6

NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURHUP WIFJGHT INITIAL

YEAR BIN ASSEMBLIES (MWd/MTIHM) (M'IIHM) ENRICHMENTii ii i i ii , ii , i Ill ,I • ,

2015 10001-15000 188 14900 35.2 3.5292015 15001-20000 176 16850 31.2 3.94,220 15 25001-30000 256 29811 47.2 3.7832015 30001-35000 151 31837 26.8 3.9012015 35001.,40000 111 37691 20.0 3.4542015 40001.45000 1691 42880 302.4 3.74620 15 45001-50080 20 45543 3.6 3.675

2016 5001-10000 16 9635 2.9 3.0852016 10001-15000 400 14660 73.6 3.2372016 15001-20000 16 19064 2.9 3.1912016 25001-30000 416 28217 76.6 3.3852016 30001-35000 67 33015 12.2 3.-9912016 3500144X_ 54,4 38708 98.8 3.4302016 40001.45000 1357 43210 243.8 3.67720 16 45001-50000 190 45180 33.5 3.915

2017 30001-35000 46 34166 8.1 3.3642017 35001-40(X_ 337 38795 61.3 3.4112017 443001.45000 918 42736 164.4 3.590

2018 15001-20000 196 15558 36.4 3,4222018 30001-35000 196 31284 36.4 3.6052018 35001..400(30 213 38821 38.1 3.5392018 443001.45000 1064 43255 192.4 3.62620 18 45001-50000 536 45218 94.2 3.843

2019 30001-35000 46 33844 8.2 3.3432019 35001-40000 233 38041 42.7 3.34920 19 40001.45000 1053 42820 186.2 3.663

2020 30001-35000 47 33104 8.4 3.2982020 35001,.40000 314 37812 56.5 3.3782020 443001.45000 1434 42271 256.6 3.5942020 45081-50000 203 45164 36.1 3.642

2021 30001-35000 3 31524 0.6 2.0452021 35001-g0(_ 304 38274 53.6 3.3432021 443001.45000 818 42233 144.6 3.6032021 45001-50000 145 45017 25.7 3.632

2022 10(30I-15000 471 13200 83.7 3.274

2022 20001-25000 116 21183 20.0 3.5552022 25001-30000 354 2804 1 63.5 3.3332022 30001-35000 164 31913 28.6 3.52.52022 35001-4(X)_ 444 37941 79.7 3.3682022 40001-45000 1496 42142 267.2 3.591

B-9

Tae4eE;2. (eontinued)i I • ii II I I II I IIII II I I I II I I III II I I I I I III

LWR OUAN'ITr'IF_ DATABASE

ProjectedData: No New Orden Casewith Era.endedBumupData Broken Down By:. D_lutrge Year and Bumup BinProjectedAssem0Uesfor Assembly Class: GE BWPd4.6

NUMBER AVERAGE TOTAL AVERAGEDISCHARGE BURNUP OF BURNUP WEIGHT INITIAL

YEAR BIN ASSEMBLIES (MWd/MTIHM) (MTIHM) ENRICHMENTI I l -- l J l l l l l l Ill Ill Ill E l

2023 X}001-35000 82 32669 14.6 3.2462023 35001-40000 821 38459 148.8 3.4682023 40001-45000 677 43486 120.9 3.742

2024 10001-15000 423 12745 76.0 2.9452024 15001.20000 656 16928 114.8 3.2632024 20001.25000 144 24913 25.4 2.8312024 25001.30000 486 27659 S6.3 3.3782024 30001.35000 7 32928 1.3 2.8572024 35001.40000 564 38382 98.5 3.2552024 40001-45000 1111 41455 196.5 3.506

2025 10001.15000 452 13275 81.9 3.3302025 15001.20000 292 17125 53.1 3.2262025 20001.25000 124 22132 22.0 3.2862025 25001.30000 553 27089 100.0 3.3252025 30001-35000 511 33895 92.5 3.2462025 35001.40000 10,,6 37165 189.2 3.3242025 40001-45f)00 136 41344 24.8 3.427

2026 10001.15000 368 13219 66.5 3.4302026 15001.20000 88 15654 16.3 3.3862026 20001.25000 252 23731 45.0 3.451:2026 25001.30000 204 27475 37.9 3.386202.6 30001.35000 263 32853 47.2 3.3932026 35001-40000 450 37803 81.8 3.4172026 40001-45000 305 41464, 53.0 3.375

2027 10001-15000 172 11160 31.8 3.6492027 15001-20000 276 17827 44.8 3.8022027 20001-25000 172 23296 31.8 3.7372027 30001.35000 287 33351 48.8 3.8982027 35001.-40000 229 36991 42.3 3.8082027 40001.45000 323 42790 55.3 3.7592027 45001-50000 40 48560 7.3 3.940

2028 30001-35000 I 19 34508 21.0 3.2852028 35001-40000 164 39737 29.1 3.299

2029 I0001-15000 332 14537 58.8 3.2102029 20001-25000 216 24280 38.3 3.2102029 30001-35000 216 32858 38.3 3.210

Grand Total 97616 _ 17503.7 3304i i

. B-10

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B-14

APPENDIXC

T2P2z

A Computer Program for Estimating

Tritium Target Performance and

Tritium Environmental Source Terms

by

Peter C. Owczarski

John R. Honekamp

January 1994

Science Applications International Corporation, Inc.Richland, WA 99352

for

Plutonium Disposition Study

under

General Electric Contract 190-PUC-9027

TABLE OF CONTENTS

1.0 INTRODUCTION AND SUMMARY ..................... I

2.0 TECHNICAL BASES .........................- 22.1 Thermal Analysis ...................... 22.2 Pressure/Stress Analysis . _ . ............... 32.3 Normal Target Permeation Leakage .............. 52.4 Tritium Leakage from Failed Target Rods ........... 5

2.5 Distribution in the Reactor System ............. 8

3.0 INPUT REQUIREMENTS ................... 9

3.1 Input File T2P2-1.DAT _ ....... •..... 93.2 Input File T2P2-2.DAT .............. 10

4.0 OUTPUT DESCRIPTION ....................... 11

5.0 REFERENCES ............................ 17

6.0 T2P2 PROGRAM LIST ......................... 18

- iii

2 • 0 I'NTRODUCTION AND SUMMARY

The purpose of this document is to descr£be both the operation and

technical bases of the T2P2 code. The code itself is a PC operated composite

collection of four codes that were written independently. The four codes were

all part of the technical procedure for estimating the performance of the

tritium target in the ABWR and estimating the tritium environmental source

terms. With the four pieces together along with additional supporting

subroutines the user can now easily optimize the target rod design while being

aware of the environmental consequences of the design or irradiation history.

In T2P2 the average target rod or pin represents the whole core load of

pins. Thus the environmental impact is from the whole core. However, peak or

low irradiation pins can be studied independently of the other pins. In the

code the chosen design and irradiation parameters are used to calculate the

pin temperature profile, internal pressure and stresses, and normal permeation

leakage. The contribution of normal leakage and failed pin leakage is modeled

in a distribution of tritium in the coolant, pool, and in environmental

streams and reactor and turbine building atmospheres.

The model for leakage of failed pins is new. In this new model the

beginning-of-life failure mode was replaced by a more realistic and

mechanistic failure mode during irradiation. This new model has reduced the

estimate of one-cycle tritium operation environmental source terms by an order

of magnitude.

This document begins with a brief description of the technical bases for

the code in Section 2. Section 3 describes the input requirements for the PC

based code. Section 4 describes the seven output files of the code. Sections

5 and 6 are the references and source code listing, respectively.

To use T2P2 for target optimization, the user should refer to

constraints on the target design discussed by Weber (1992) and Lanning et al.

(1992). These constraints along with those necessary to the ABWR and

production goals should compose a fairly complete basis for initial target

design. Eventually any final pin design would require a T2P2 upgrade that can

examine pin behavior along its axis as well as its position in the reactor.

2.0 TECHNICAL BASES

This section provides brief discussions of the technical bases for each

of the major models in T2P2. The user is encouraged to refer to the source

code listing in this document to examine the calculational methods used and

the expressions used to represent the technical bases.

T2P2 is a collectlon of codes and subroutines. Four codes were written

and used independently, but they depended on each other via input/output

files. Combining them dramatically improved the efficiency of using them and

made parametric studies much easier. The original codes consisted of a

thermal analysis of the target pin at power, an analysis of the reactor

coolant/refueling pool tritium inventory, calculations of pin pressure/stress,

and computation of normal permeation leakage of tritium from the pin. The

first two were FORTRAN codes, the third was a spreadsheet program, and the

fourth was written in MATHEMATICA. The combined program is now in FORTRAN and

includes supporting subroutines. Five crucial parts to T2P2 are explained

more fully below.

T2P2 was programmed to apply to the average target pin performance.

Although the peak irradiation rod performance can be examined on an individual

basis, the overall response of the tritium in the coolant and the

environmental source terms should be interpreted only for the average pin

which represents the contribution of all the pins in the core.

Target pins are assumed to produce tritium at a constant rate during

irradiation. The leakage of tritium to the coolant from the pins also

proceeds at a constant rate. However, the concentration of the tritium in the

coolant responds to loss rates proportional to the tritium concentration with

constant proportionality coefficients (e.g., radioactive decay) during power

and with other constant coefficients during refueling. The resulting coolant

tritium concentration exhibits nonlinear transient behavior.

2.1 Thermal Analysis

The temperature profile in the tritium target is calculated by solving

the steady state equation for conduction in a solid with heat generation for a

composite cylinder of concentric cylindrical layers with heat transport out in

the radial direction only. The added heat transport across the gas gaps due

to radiation and free convection are ignored, so the simulation produces

slightly higher temperatures than reality. The heat generation per unit

volume is calculated for each layer (see Figure 2.0) from the gamma energy

C-2

absorbed and the n-u energy (LiAIO 2 only). The user must supply the average

coolant temperature. Conductivities for the various layers were taken from

Wilson (1991).

The temperature profile is used in the calculations of the gas pressure

in the pin during irradiation, the permeability of the pin to tritium gas, and

the yield stress of the cladding.

2.2 Pressure/Stress Analysis

The technical bases for subroutine GETARG are described as follows.

This subroutine first computes the free gas space in the pin from the pin

dimensions. The gas is assumed to be helium from the initial fill at one

atmosphere and 32 °F and helium from the n-a reaction in the pellet column

which does not retain any helium. The ideal gas law with free gas space

temperature, volume and free gas moles gives the interior pressure.

The interior pressure is calculated both at operating temperature andat

IO0°F (refueling condition). The exterior pin pressure is either the water

vapor pressure at the coolant temperature (at power) or 30 feet of water head

(at refueling). The interior-exterior pressure differences are used to

C-3

Center Gas Space

Zircaloy Liner

LiAIO 2 Target

Gap

Nickel Plated Zircaloy Getter

Gap

Aluminide Barrier

Stainless Steel Cladding

Figure 2.0. Reference Target Pin Radial Cross Section.(to scale with cladding o.d. = 0.483 in.)

calculate the pin cladding hoop and axial stresses at power and refueling and

compared with the 90% yield stress of the stainless steel cladding.

Other useful calculations performed by GETARG are the GVR and the total

hydrogen/getter zirconium ratio (TH/Zr). The GVR is the ratio of STP gas

volume produced to the original pellet volume. The TH/Zr is 1.1 times the

atoms of tritlum/atoms getter zirconium. The 1.1 factor (based on In-reactor

test results} reflects the likelihood that some protium from the primary

coolant diffuses into the pin.

GETARG also supplies the heat generation rates for the thermal analysis.

2.3 Normal Target Permeation Leakage

No materials are now known that can be placed in the target

configuration and under reactor conditions can totally prevent any tritium

from escaping the pin. Since all the materials of construction have some non-

zero permeability, it is necessary to be able to estimate the permeation

losses from the pin in terms of the internal tritium partial pressure and the

barrier permeabilities. The portion of the main T2P2 code called TLEAK

estimates this permeation loss as a function of irradiation time. Sherwood

(1992) gives a detailed description of the models within TLEAK; the

descriptions are not repeated here. The models are empirical representations

of physical data that include the solubility of tritium gas in the pellet and

getter and the permeation rate through the cladding. The permeation leakage

depends on the temperature of the pin components as well as the pin

dimensions. Any studies made with T2P2 would be sensitive to changes in the

permeation model parameters.

2.4 Tritium Leakage from Failed Target Rods

None of the getter-barrier target rods irradiated during the DOE Tritium

Target Development Program developed cladding leaks. Thus, there are no

directly applicable experimental data on which to base clad failure

predictions or tritium leakage from a failed target rod. However, there are

extensive data for Zircaloy clad LWR fuel elements and Stainless Steel clad

fast reactor fuel elements.

The prior model used to estimate tritium target rod failures (Apley

1992) was based on the premise that the target rod cladding was less likely to

fail than the LWR fuel rod cladding which operated under more stringent

conditions. Post-irradiation examination of the getter-barrier target rod

C _ 5

(Lanning 1992) has confirmed that its cladding did not experience any chemical

or mechanical interaction with its internal components and could be modeled as

a free-standing gas-pressurized tube. On the other hand, LWR fuel element

cladding is subject to internal corrosion, hydrldlng and fuel-clad mechanical

interaction in addition to gas pressure. Further, the 20% cold-worked 316-

stainless steel used for the target rod is a much stronger material than

zircaloy at equivalent temperatures.

The Apley model used a clad failure frequency of 10_ased on LWR fuel

element experience and assumed that 50% of the tritium inventory in the failed

rod would be released. The 50% release implies that all cladding failures

exist from beginning of life and that the capacity for retaining tritium is

determined by the solubility of tritium in the LiA102. This assumption was

considered to be very conservative, since most failures would likely occur

late in life when essentially all the tritium is tied up in the zirconium

getter and pellet. Further, throughout the target rod life very little

tritium is present in gas phase. However, because the tritium permeation from

intact target rods is so low, the release from failed target rods using the

Apley model dominates the tritium source term.

To overcome this limitation a more mechanistic and realistic target rod

failure model was developed for T2P2. First we assume a cladding failure mode

similar to stainless steel clad LMR fuel elements. These failures are

typically microscopic pin-holes that can take many hours to depressurize. We

also assume that both the LiAIO2 pellets and the zircaloy getter tubes will

retain their tritium inventory at the time of clad failure. This assumption

is based on the knowledge that both the pellets and getters are stable in

water and steam at the ABWR operating conditions.

The model assumes that there ks no steam or water ingress into a target

rod with a pin-hole cladding failure until a shutdown - restart cycle occurs.

This is based on the premise that clad failure is very unlikely until the

internal helium pressure exceeds the primary system pressure. Further, as

long as the internal helium pressure is equal to or greater than the primary

system pressure, and the hole in the cladding remains small, the getter would

continue to perform its function. However, during the first shutdown -

restart cycle following clad failure it is assumed that the target rod

depressurizes and becomes water/steam logged. After the shutdown - restart

cycle we assume that the getter is oxidized, ceases to function as an internal

tritium sink, and all tritium released from the LiAIO2 escapes to the coolant.

The amount of tritium leaving the pellet is governed by tritium

solubility. This solubility is designated GVRo (Sherwood 1992}. GVRo

decreases as temperature increases and ks about 40-60. For GVR < GVRo, only

3% of the tritium formed escapes the pellet due to recoil. For GVR > GVRo

100% of tritium formed escapes the pellet. Thus we have defined that the

availability of tritium to leave the pan depends on the degree of irradiation

and the ingress of water into the pin after a shutdown cycle followlng the

tim8 of failure.

In Subroutine FAIL we have assigned a frequency of shutdown cycles to

two (the user can recompile T2P2 with a different number} with the usual

refueling periods. That leaves three periods per refueling cycle in which the

pins can fail. Each period has a different consequence of tritium leakage.

We now explain the likelihood of failure.

Target rods failures are assumed to follow a Weibull distribution

similar to stainless steel clad LMR fuel elements with a failure frequency of

10 4 over a three year irradiation period. The Weibull distribution function,

and the details for selecting the three parameters involved, are documented in

the references listed in the source code for Subroutine FAIL. The model

quantifies the failure assumptions and produces a tritium source term to the

coolant for the exposure cycle between target pin loadings. The source term

rate is assumed to be linear over the loading cycle and is added directly to

the normal permeation leakage term.

The overall effect of the new model is to delay the failure to an

exposure determined by the Weibull distribution function and the frequency of

shutdown-restart cycles. This reduces the impact of failure because only

tritium released from the LiAIO 2 after clad failure escapes to the coolant.

For the parameters selected, this model reduced the consequences of failure by

an order of magnitude compared to the Apley model. Further, with this model

the tritium release is readily adjusted to accommodate actual operating

experience with respect to shutdown-restart cycles and clad failure

statistics.

C A 7

2.5 Distribution in the Reactor Systea

The distribution of tritium in the reactor coolant system follows a

simple first order differential equation:

Dc/dt = S - (L l + L2)C

where C = bulk coolant curies of tritium

t = time

S = source leakage rate from pins

Li = fractional rate of loss of coolant via steam leaks

L2 = fractional rate of loss of coolant via pool evaporation.

When pool liquid is isolated from the rest of the coolant during power,

a separate differential equation with only a L2 term is solved for that

liquid.

The Ls and S are held constant during power and with Ll=S=O during

refueling. The Subroutine H3CYCLE listing gives the details of the solution

to the differential equations.

The model for pool evaporation contains an evaporation mass transfer

model and assumes instantaneous mixing of the evaporation source with reactor

building air giving a bulk air concentration of tritium in the building. This

simple model has room for upgrading in two areas. The first might be to

establish local or 3-d concentration profiles, if possible, for detailed

worker exposure. The second is to provide a mechanistic mass transfer

coefficient based on local air flow velocities, rather than use a recommended

but nonmechanistic default value.

C^ 8

3.0 INPUT REQUI_4m_TS

There are two separate input files needed to run T2P2: T2P2-1.DAT and

T2P2-2.DAT. The latter file input varlables control the disposition of

tritium throughout the reactor system and control the environmental source

term. The former file variables control the target pin behavior in the

reactor. The simplest way to describe the file variables Is to use an example

run throughout this document. A description of each number used in the

example is found below. An important distinction in input values in the two

DAT files is that the duration of the refueling cycle does not have to be the

same as the tritium target irradiation cycle.

3.1 Input File T2P2-I.DAT

The example _ile consists of the following five rows of numbers:

9,538.

.5893,.6096,.9906,1.0084,1.049,1.0592,1.0744,1.226816.3,3.18,.241,16.3,.241,100.,18.22.866E+13,381.,25.4,2.15,.5,273.75,365.

10,0.,1.,30.,60.,90.,120.,150.,180.,210.,240.,273.75

9 = number of concentric regions in the pin, not counting thecenter hole.

538. = average coolant temperature, degrees F.

.5893 = inner diameter of liner, cm.

.6096 = outer diameter of liner = inner diameter of pellet, cm. etc.to 1.2268 = outer cladding diameter.

16.3---18.2 = thermal conductivity of the 7 regions, watts/m/_.

2.866E+13 = n-alpha reaction rate, rxns/s/cm pellet.

381. = pellet column length, cm.

25.4 = pin plenum length, cm.

2.15 = gamma heating rate, watts/g.

.5 = lithium enrichment, fraction of Li-6 at beginning of life.

273.75 = duration of tritium irradiaton/cycle, days.

365. = duration of cycle, including refuelling, days.

i0 = number of time steps desired for normal pin leakagecalculations.

0.---273.75 = I0 time points, in days, for leakage calculations.

The input numbers have to be placed in the five rows. If this cannot be done,

then the code READ statements will have to be changed for the too long rows.

3.2 Input File T2P2-2.DAT

The example file consists of the following eight rowsz

9,273.75,91.25,273.75,91.254.86e+6,10.52e+63488.

83000.,35000.,201000.1,1,1

9 ffi number of consecutive cycles.

273.75 = number of days at full power in one refueling cycle.

91.25 = number of days off power in one cycle.

273.75 = differential days at power for calculational points =2_3.75/I

91.25 = differential days off power for calculational points =91.25/1

4.86E+6 = reactor coolant inventory at powers ibm.

10.52E+6 = additional refueling pool inventory, ibm.

3488. = number of target rods in core.

1 = number of leak paths from coolant during power.

1 = number of leak paths from pool during power.

1 = number of leak paths from pool and coolant off power.

C- I0

4.0 OUTPUT DESCRIPTION

The output from T2P2 consists of seven files. These files provide the

user with tritium target rod performance, reactor tritium inventories, and

environmental source terms. Each of the file numbers and titles are listed

below.

List of output files:

T2P2-1.OUT TARGET THERMAL ANALYSIS OUTPUTT2P2-2.OUT NORMAL TARGET LEAKAGE ANALYSIS OUTPUT

T2P2-3.OUT GVR/PRESSURE/STRESS CALCULATIONST2P2-4.OUT REFUELING CYCLE COOLANT INVENTORYT2P2-5.OUT LEAKAGE PARAMETERS & OUTPUTT2P2-6.OUT TRITIUM Ci BALANCE CHECKS

T2P2-7.OUT TRITIUM DISPOSITION OUTSIDE REACTOR

Sample output files follow.

T2P2-1.OUT TARGET THERMAL ANALYSIS OUTPUT

Description: This output file provides the steady state temperatureprofile of the target pin. The concentric layers have outer radii,r(out), in meters, thermal conductivities in w/m/K, heat generation

rates in w/m3, outer radii temperatures, and volume average temperaturesin degrees Celcius.

layer r(out) t cond ht sen puv degrees C

1 3.048000E-03 16.300000 1.443254E+07 359.6986002 4.953000E-03 3.180000 5.172376E+07 333.7321003 5.042000E-03 2.410000E-01 0.O00000E+00 304.2784004 5.245000E-03 16.300000 1.443254E+07 303.295400

5 .296000E-03 2.410000E-01 0.000000E+00 286.6868006 5.372000E-03 100.000000 1.702800E+07 286.6278007 6.134000E-03 18.200000 1.702800E+07 283.302700

layer Tavg

1 359.7016002 349.5846003 318.9178004 303.783600

5 294.9643006 286.657400

7 284.937100

TGASR 1126.809000 des R gas temp fcr press calcs

T2P2-2.OUT NORMAL TARGET LEAKAGE ANALYSIS OUTPUT

Definitions:

tday = days irradiation time at end of periodrl = tritium permeation rate, gmoles/secondnt2 = tritium gram moles in getter

totalt2 = total gram moles tritium produced

C° ii

iperm - target permeation rate, gmoles T2/sec

relCi - g-moles T2 released in time periodsumCl - cumulative Tritium Ci released to coolant

p - tritium gas pressure, Pascalstzratlo - ratio of T atoms/Zr atomslkratlo - ratio of leaked to total tritium

totCi - T Ci inventory in target after decay

tday rl nt2 totalt2

1.000000 2.379213E-11 7.831382E-06 7.831382E-0430.000000 5.078669E-10 2.606835E-03 2.344148E-0260.000000 9.073245E-10 9.634805E-03 4.677434E-0290.000000 1.221514E-09 2.011289E-02 6.999906E-02

120.000000 1.468536E-09 3.327537E-02 9.311616E-02

150.000000 1.663008E-09 4.852151E-02 1.161261E-01180.000000 1.815889E-09 6.537999E-02 1.390295E-01210.000000 1.936136E-09 8.348110E-02 1.618267E-01240.000000 2.030716E-09 1.025348E-01 1.845182E-01273.750000 2.113211E-09 1.248276E-01 2.099206E-01

iperm relCi sumCi p

5.076373E-15 2.192993E-10 1.276174E-05 1.805498E-042.953369E-14 4.335949E-08 2.535991E-03 6.111185E-034.254178E-14 9.340981E-08 7.971810E-03 1.268005E-024.985401E-14 1.197449E-07 1.494016E-02 1.741366E-025.508119E-14 1.359960E-07 2.285420E-02 2.125673E-025.894406E-14 1.477767E-07 3.145381E-02 2.434277E-026.185364E-14 1.565538E-07 4.056418E-02 2.680528E-026.407187E-14 1.631994E-07 5.006129E-02 2.876237E-026_577690E-14 1.682840E-07 5.985428E-02 3.031354E-02

6.723746E-14 1.939349E-07 7.113998E-02 3.167469E-02

tzratio Ikratio totCi

8.807521E-06 2.779307E-07 45.9169702.931762E-03 1.845134E-06 1374.4210001.083573E-02 2.906796E-06 2742.4740002.261985E-02 3.640224E-06 4104.1860003.742297E-02 4.186067E-06 5459.5880005.456945E-02 4.619644E-06 6808.7090007.352925E-02 4.976236E-06 8151.5790009.388657E-02 5.276149E-06 9488.225000

1.153152E-01 5.532495E-06 10818.6800001.403867E-01 5.779947E-06 12308.070000

T2P2-3.OUT GVR/PRESSURE/STRESS CALCULATIONS

This output focuses on the end of the target irradiation period (eol)in the reactor. At this point the interior pan pressure is at maximumwhile still at power with the coolant at maximum temperature. Thecoolant ks then cooled to IOOF. Symbol definitions:

VVR = void volume/pellet volume90%YS = 90% yield stressTH/Zr ratio = atoms of H+T/atoms Zr

Vcladid 358.095800 Vpellet 181.525100 cm**3

. C-12

Vgetter 24.870410 Vget Ni 21.758550

Vllner 7.246315 VsprTng 3.070000Vnet 141.383900 VVR 7.788669E-01GVR 79.420030

GVRo 43.311820

Li6 depletion % 12.024320

Reactor Cycle Capacity Factor 75.000000 %

T2 molecules, no decay 1.291335E+23T2 molecules, w/ decay 1.264352E+23T2 curies 12308.070000 TH/Zr ratio 3.025226E-01Fraction of T2 in gas 4.012920E-09

heat generation watts/cm**3:cladding 17.028000getter 14.432540liner 13.760000

pellet 51.723760

internal pressure at power 2323.939000 psiinternal pressure at 100 F 1154.948000 psi

hoop stress at power 9288.527000 psiaxial stress at power 4644.264000 psi

hoop stress at 100 F 7688.197000 psiaxial stress at 100 F 3844.099000 psi

90%YS@pwr 71737.800000 90%YS100F 82296.000000 psi

netDP@pwr 1361.992000 netDPIOOF 1127.333000 psi

Puff T2 Ci released from sudden eol pin failure:2.894677E-05 at power to coolant4.821030E-05 @ IOOF in pool4.873327E-05 @ 75F in air at 1 arm

T2P2-4.OUT REFUELING CYCLE COOLANT INVENTORY

Input Parameters:

Number of cycles 9Days at full power 273.750000Days of refueling 91.250000Lbm coolant 4860000.000000

Lbm refuel pool 1.052000E+07

Days dt at power 273.750000Days dt refueling 91.250000

CYCLE NUMBER ( i)

Days Coolant H3, Ci Pool H3, Ci

273.750000 103.991000 0.000000E+00 at power

365.000000 27.674740 59.904990 refueling

CYCLE NUMBER ( 2)

Days Coolant H3, Ci Pool H3, Ci

_- 13

273.750000 104.568100 38.834530 at power365.000000 38.163210 82.608430 refueling

CYCLE NUMBER ( 3)

Days Coolant H3, Ci Pool H3, Ci273.750000 104.786800 53.552460 at power

365.000000 42.138250 91.212840 refueling

CYCLES 4-8 OMITTED BELOW.

CYCLE NUMBER ( 9)

Days Coolant H3, Ci Pool H3, Ci273.750000 104.919900 62.507940 at power365.000000 44.556960 96.448390 refueling

T2P2-5.OUT LEAKAGE PARAMETERS & OUTPUT

Discussion: This output consists of cumulative and cycle incremental tritiumairborne leaks. Here path one is the turbine building stack and path 2 is the

reactor building stack. Loop leakage at power is from the turbine buildingsteam loss. Pool leakage at power is from evaporation of refueling pool waterduring power. Pl+ip leakage during refueling is also pool evaporation, whichincludes coolant water at that time.

Leakage parameters:

No. of target rods 3488.000000Frac. Fail T2 Loss 3.794628E-06

Ci leaked/cycle 411.041700H3 Ci/day source 1.501522

loop leak rate no. 1 at power 67962.000000 ibm/daypool leak rate no. 1 at power 15038.000000 lbm/dayip+pl leak rate no. 1 refueling 26581.730000 lbm/day

CYCLE NUMBER ( 1) ************************

cycle day= 273.750000 (at power)Total loop leak from path 1 303.708000 CiTotal pool leak from path 1 0.000000E+00 Ci

cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 15.069300 Ci

Cycle 1 Incremental Leak Values:

Cycle loop leak from path 1 303.708000 CiCycle pool leak from path 1 0.O00000E+00 CiCycle pl+ip leak from path 1 15.069300 Ci

CYCLE NUMBER ( 2) ************************

cycle day= 273.750000 (at power)Total loop leak from path i 634.218600 Ci

([-, 14

Total pool leak from path 1 19.022310 Ci

cycle day- 365.000000 (refueling)Total pl+Ip leak from path 1 35.849730 Ci

Cycle 2 Incremental Leak Values:

Cycle loop leak from path 1 330.510600 CiCycle pool leak from path 1 19.022310 CiCycle pl+lp leak from path 1 20.780430 Ci

CYCLE NUMBER ( 3) ************************

cycle day= 273.75C000 (at power)Total loop leak from path 1 974.887100 CiTotal pool leak from path 1 45.253910 Ci

cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 58.794630 Ci

Cycle 3 Incremental Leak Values:

Cycle loop leak from path 1 340.668500 CiCycle pool leak from path 1 26.231600 CiCycle pl+_p leak from path 1 22.944900 Ci

CYCLES 4-8 OMITTED BELOW.

CYCLE NUMBER ( 9) ************************

cycle day= 273.750000 (at power)Total loop leak from path 1 3052.322000 CiTotal pool leak from path 1 226.364600 Ci

cycle day= 365.000000 (refueling)Total pl+ip leak from path 1 203.585900 Ci

Cycle 9 Incremental Leak Values:

Cycle loop leak from path 1 346.849400 CiCycle pool leak from path 1 30.618260 CiCycle pl+ip leak from path 1 24.261920 Ci

T2P2-6.OUT TRITIUM Ci BALANCE CHECKS

cycle number 1source total 411.041700

coolant inventory 27.674740

refuel pool inv y 59.904990Ci leaked 318.777300

Ci decayed 4.684668net difference -2.336502E-05

cycle number 2source total 822.083300

coolant inventory 38.163210

refuel pool inv_y 82.60B430Ci leaked 689.090600

Ci decayed 12.221070net difference -3.433228E-05

C- 15

cycle number 3source total 1233.125000

coolant inventory 42.138250

refuel pool inv_y 91.212840Ci leaked 1078.936000

Ci decayed 20.838260net difference -1.525879E-05

CYCLES 4-9 OMITTED.

T2P2-7.OUT TRITIUM DISPOSITION OUTSIDE REACTOR

The values printed below can be used to estimate offsite doses (from buildinglosses) and occupational doses (from building airborne concentrations). Dosesmust be computed for tritium as part of a water molecule (TOH or T20), not asT2 or HT. If tritium is the only contributor to occupational dose, then alimit of 2E-06 microCi/cc can be tolerated before respiratory protection isrequired.

The following Ci and concentrations are for the two periods in the365.000000 day production cycle number 9 with

273.750000 days at power and 91.250000 days refueling.

Reactor Building:

Total evap. loss during power = 30.618260 CiAir H3 conc during power = 7.837018E-08 microCi/cc

Total evap. loss refueling = 24.261920 CiAir H3 conc refueling = 1.863017E-07 microCi/cc

Turbine Building:

Total steam loss during power = 346.849400 CiAir H3 conc during power = 1.545907E-07 microCi/cc

Other Losses:

No other losses.

C- 16

5.0 REFERENCES

Apley, W.J. 1992. Tritlum Tarqet Development Pro_ect Executive Summar 7

Topical Report. PNL-8142, Pacific Northwest Laboratory, Richland, WA.

Lanning, D.D., D.L. Baldwin, and R.J. Guenther. 1992. Final Report on theWC-1Liqht-Water Reactor Tarqet Rod Irradiation Test and postirrQdiation

Examinations. PNL-8133 Volume 1: Text, Pacific Northwest Laboratory, Richland,WA.

Sherwood, D.J. 1992. Modelinq the Behavior of a Liqht-Water Production_eactor Tarqet Rod. PNL-8010, Pacific Northwest Laboratory, Richland, WA.

Weber, J.W. 1992. Topical Report: NPLWR Tritium Tarqet Desiqn. WHC-SP-0840,Westinghouse Hanford Company, Richland, WA.

Wilson, D.R. 1991. TKTARI: A Computer Code for Predictinq Tritium Tarqet RodPerformance. WHC-SP-0684, Westinghouse Hanford Company, Richland, WA.

C_17

6.0 T2P2 PROGRAM LIST

C Program T2P2 (Tritium Target Performance Program) combines TATT, TLEAK,C GETARG, H3CYCLE,and EVAP and now employs user inputs for parametricC studies of the controlling variables.

C ************************************************************************

C Program TATT (Thermal Analysis Tritium Target) performs the steady stateC temperature analysis of a tritium target rod.

REAL k(lO), A(10), d(0:ll),C(2,0:ll),Tr(0:11),Trf(O:ll),Trc(0:11),+Trav(10),r(0:11)REAL len, mzr,molzr,mpr,ka,tday(O:lO),tsec(O:lO),rl(lO),nt2(lO),

+totlt2(lO),pus(10),pe(10),p(10),iperm(10),relCi(10),sumCi(10),+ipave,tzr(10),lambdaT,n_A

COMMON/BLK1/Nt,A,d, len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLKI0/GVR,GVRoCOMMON/BLK2/sumc,totCiCOMMON/BLK5/tfluid,tcladf

Open(Unit=l,File='t2p2-1.dat',status='old')Open(Unit=2,File='t2p2-1.out',status='unknown')

C SI units used in calculations.

C k(j) thermal conductivity of concentric region j.C r(j) outer radius of region j.C C(m,j) integration constants - to be determined.C A(j) heat generation puv in region j.

C Nt = no. of concentric regions less the center gas core.

C Tf = coolant temperature, K.C h = cladding-to-coolant heat transfer coefficient.

Read(l,*)Nt,TfluidFTf= (TfluidF+460.)/l.8h=40000.lambdaT=l.792E-09

Read(l,*)(_i(j),j=0,Nt)do l=0,Nt

r(1)=d(1)/200.enddo

Read(l,*)(k(j),j=l,Nt)

Read(l,*)n A, len,plen,gammaen,enrichli,FPcycle,EOcycle

c n A = n-alpha rxn rate per cm pellet/secondc fen = pellet length, cmc plen = rod plenum length, cm

ntgr=lcall getarg

do m=l,Nt

A(m)=I.E+6*A(m)

C.-18

enddo

c Define c(m,n) matrix:

C(1,1)= A(1)*r(0)**2./2./k(1)DO J=l,Nt-1

C(l,J+l)=r(J}*((A(J+l)-A(J))*r(J)/2.+C(l,J)*k(J}/r(J))/k(J+l)ENDDO

c Cladding surface temperature Trn

Trn=Tf-(C(l,Nt}*k(Nt}/r(Nt}-A(Nt)*r(Nt}/2.)/h

C(2,Nt)=Trn+A(Nt)*r(Nt)**2./4./k(Nt)-C(l,Nt)*ALOG(r(Nt))

DO J=Nt-l,1,-i

C(2,J)=C(2,J+I)+(C(1,J+I)-C(I,J))*ALOG(r(J)}++ (A(J)/k(J)-A(J+l)/k(J+l})*(r(J}**2.)/4.ENDDO

C(2,0)=-A(1)*r(0)**2./4./k(1)+C(I,I)*ALOG(r(O))+C(2,1)Tr(O)=C(2,0)

write(2,*)' '

write(2,*)' 'write(2,*)' TARGET THERMAL ANALYSIS OUTPUT'write(2,*)' 'write(2,*)' '

write(2,*)'Description: This output file provides the steady stat+e temperature"

write(2,*)'profile of the target pin. The concentric layers have+outer radii,'

write(2,*)'r(out), in meters, thermal conductivities in w/m/K, hea+t generation'write(2,*)'rates in w/m3, outer radii temperatures, and volume ave

+rage temperatures'

write(2,*)'in degrees Celcius.'write(2,*)' '

write(2,*)' 'do j=l,Nt

Tr(J)=-A(J)*r(J)**2./4./k(J)+C(I,J)*ALOG(r(J))+C(2,J)tav=-A(J)*(r(J)**4.-r(J-l)**4.)/8./k(J)+(C(2,J)-C(l,J)/2.)*

+ (r(J)**2.-r(J-l)**2.)++ C(l,J)*(alog(r(j))*r(J)**2.-alog(r(J-l))*r(J-l)**2.)

Trav(J)=tav/(r(J)**2.-r(J-l)**2.) - 273.16enddo

Tpellet=Trav(2)+273.16

do m=0,Nt

Trf(m)=Tr(m)*l.8-460.Trc(m)=Tr(m)-273.16

c write(2,*) m,Trf(m),' F',Tr(m),' K',Trc(m),' C'enddo

write(2,*)' layer r(out) t cond ht gen+puv degrees C'write(2,*)' 'do m=l,Nt

write(2,*)m,r(m),k(m),A(m),Trc(m)enddo

write(2,*)' '

C" 19

write(2,*)' layer Tavg'write(2,*)' 'do n=l,Nt

write(2,*} n , Tray(n)enddo

tcladf=Trf(Nt)tfluid=tfluidf

vgsinr=r(0)**2.

vgsgapl=r(3)**2.-r(2)**2.vgsgap2=r(5)**2.-r(4)**2.vgst=vgsinr+vgsgapl+vgsgap2tnr=Tr(0)-273.16

tgsc=(vgsinr*tnr+vgsgapl*Trav(3)+vgsgap2*Trav(5))/vgstTGASR=tgsc*l.8+492.write(2,*)' '

write(2,*)' TGASR ',TGASR,' deg R gas temp for press calcs'ntgr-2

c call getarg

C ************************************************************************

C Program TLEAK computes the permeation loss of a tritium target rodC during irradiation in the manner of EP Simonen using a MathematicaC language code (see 4/9/93 transmittal).

Open(Unit=3,File='t2p2-2.out',status='unknown,)Read(1,*)mt,(tday(j),j=0,mt)

pi=3.14159rgas=8.314len=381.

rpin=r(1)*100.rpout=r(2)*100.rgin=r(3)

rgout=r(4)rci=r(5)rco=r(7)a1=2.*pi*(rgin+rgout)

vl=pi*(rpout**2.-rpin**2.)vg=pi*(rgout**2.-rgin**2.)*len*10000.rzr=6.49

mzr=rzr*vgamuzr=91.22

molzr=mzr/amuzrwclpav=8.87E+13

pavg=n_A/vl

tc=(Trav(6)*(r(6)**2.-r(5)**2.)+Trav(7),(r(7)**2.-r(6)**2.))/+(r(7)**2.-r(5)**2.)tg=Trav(4)tp=Trav(2)

do j=O,mt

tsec(j)=tday(j)*86400.enddo

C" 20

eol=tsec(mt)mpr=pavg*vl*lO0./6.023E+23

taudif=6.65E-lO*exp(l.31/(8.62E-5*(tp+273.)))tausur=66.4*wclpav/pavgreltau-taudif+tausur

qa=8156.,4.184ka-4.1*exp(-qa/(rgas*(tg+273.)))xp=-65700./(rgas*(tc+273.))

pss-1.01E+5**(-.5)*2.33E-2*exp(xp)/224.14qe=36910.,4.184ps=638758.6*exp(-qe/(rgas*(tg+273.)))*760_*133.3

c Time marching sequence 0<t<eol.

sumc=O.

DO j=l,mt

emprl=O.5*mpr*(l.-exp(-tsec(j)/(86400.*reltau)))rl(j)=O.5*0.01*mprif(rl(j).it.emprl)rl(j)=emprl

ts=tsec(j)call INGRATE(reltau,ts,rint)

nt2(j)=0.5*mpr*rint*len/100.

totlt2(j)=mpr*tsec(j)*len/200.

c Correct nt2 and totlt2 for tritium radioactive decay. Correctionc is based on constant production rate.

xdk=ts*lambdaT

cordk=(l.-exp(-xdk))/xdknt2(j)=cordk*nt2(j)totlt2(j)=cordk*totlt2(j)

zp=-9.632E+4/(rgas*(tg+273.))pus(j)=2.888*(2.*nt2(j)*123.2/molzr)**2.*760.*133.3*exp(zp)pe(j)=pus(j)if(pus(j).gt.ps)pe(j)=ps

p(j)=(rl(j)*rgas*(tg+273.)/(al*ka)+pe(j)**.875)**(l./.875)prfave=100.

iperm(j)=2.*pi*pss*(len/lOO.)*p(j)**.5/prfave/alog(rco/rci)

dt=tsec(j)-tsec(j-l)

if(j.eq.l)thenipave=iperm(j)/2.

else

ipave=(iperm(j)+iperm(j-l))/2.endif

relCi(j)=ipave*dtsumc=sumc+2.*6.023E+23*relCi(j)/2.07E+19sumCi(j)=sumctzr(j)=2.*nt2(j)/molzr

ENDDO

C-21

pT2=p(mt)

write(3,*}' '

write(3,*}' 'write(3,*)' NORMAL TARGET LEAKAGE ANALYSIS OUTPUT'

write(3,*}' 'write(3,*}' 'write(3,*}'Definitions:'

write(3,*}' tday - days irradiation time at end of period"write(3,*}' rl - tritium permeation rate, gmoles/second'write(3,*}" nt2 - tritium gram moles in getter'write(3,*}' totalt2 - total gram moles tritium produced'write(3,*}' iperm = target permeation rate, gmoles T2/sec'write(3,*)' relCi = g-moles T2 released in time period'write(3,*}' sumCi = cumulative Tritium Ci released to coolant'

write(3,*}' p = tritium gas pressure, Pascals'write(3,*)' tzratio = ratio of T atoms/Zr atoms'write(3,*)' lkratio = ratio of leaked to total tritium'

write(3,*)' totCi = T Ci inventory in target after decay'write(3,*)' 'write(3,*)' '

write(3,*}' tday rl nt2+totalt2'

write(3,*)' '

do j=l,mt

write(3,*)tday(j),rl(j),nt2(j),totlt2(j)enddo

write(3,*)' 'write(3,*)' iperm relCi sumCi

+ p'write(3,*)' '

do j=l,mt

write(3,*)iperm(j),relCi(j),sumCi(j),p(j)enddo

write(3,*)' '

write(3,*)' tzratio ikratio totCi'write(3,*)' '

c Specific Activity of T2 = 9720 Ci/gram x 6.0321gram/gmolesat2=9720.*6.0321

do j=l,mt

totCi=totlt2(j)*sat2tlkr=sumCi(j)/totlt2(j)/sat2write(3,*)tzr(j),tlkr,totCi

enddo

call getargcall RFcycle

stopend

C ************************************************************************

Subroutine RFCYCLE

C Code RFCYCLE computes losses of H3 from the coolant system of a

C - 22

C reactor for successive refueling cycles.

REAL Lr(lO),Lb(10),Cb(lO),Cr(lO},idk,Mo,Lkb,Lkr,netdif

Real Mr,Cbp(lO),Lbp(10),Lkbp,Ntarg,Edot(2),Lxtrareal dCb(lO),dCbp(10),dCr(lO)real A(lO),d(O:ll),n a

COMMON/BLK1/Nt,A,d, len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLKIO/GVR,GVRoCOMMON/BLK2/sumc,totCiCOMMON/blk3/Edot,RBacfmCOMMON/BLK4/FailT2fr,rrctrc,tb

c Lr and Lb are system leak rates in lbs/day during refueling andc buildup, respectively.c Cb and Cr are the corresponding cumulative H3 losses in Curies with noc radioactive decay once leaked.

Open(Unit=ll,File='t2p2-2.dat',status='old'}Open(Unit=12,File='t2p2-4.out',status='unknown')Open(Unit=13,File='t2p2-5.out',status='unknown'}Open(Unit=14,File='t2p2-6.out',status='unknown'}Open(Unit=15,File='t2p2-7.out',status='unknown')write(12,*)' 'write(13,*)' 'write(12,*)' ***REFUELING CYCLE COOLANT INVENTORY ***'

write(13,*)' ******LEAKAGE PARAMETERS & OUTPUT******'write(14,*)' **_****TRITIUM Ci BALANCE CHECKS*******'write(13,*)' 'WRITE(12,*)' 'Write(12,*)' Input Parameters:'write(12,*)' 'Read(ll,*)Ncm,tb,tr,dtb,dtrRead(ll,*)Mo,MrRead(11,*)NtargRead(ll,*)Wfeed,RBacfm,TBacfm

c rrctrc=ratio of tritium cycle length/refueling cycle lengthrrctrc=EOcycle/(tb+tr)

call FAIL

tlpc=Ntarg*(FailT2fr*totCi+sumc)Cdot=tlpc/tb/rrctrc

write(12,*)' Number of cycles ',Ncm

write(12,*)' Days at full power',tbwr_ e(12,*)' Days of refueling ',trwr_ue(12,*}' Lbm coolant ',Mowrite(12,*)' Lbm refuel pool ',Mrwrite(12,*)' Days dt at power ',dtbwrite(12,*)' Days dt refueling ',dtr

write(13,*)'Discussion: This output consists of cumulative and cy+cle incremental tritium'

write(13,*)'airborne leaks. Here path one is the turbine building+ stack and path 2 is the'

write(13,*)'reactor building stack. Loop leakage at power is from+ the turbine building'write(13,*)'steam loss. Pool leakage at power is from evaporation

+ of refueling pool water'

write(13,*)'during power. Pl+ip leakage during refueling is also+pool evaporation, which'

C _ 23

write(13,*)'includes coolant water at that time.'write(13,*}' 'write(13,*)' 'write(13,*)' Leakage parameters_'write(13,*)' '

write(13,*)' No. of target rods ',Ntargwrite(13,*)' Frac. Fail T2 Loss ',FailT2frwrite(13,*)' Ci leaked/cycle ",tlpcwrite(13,*}' H3 Ci/day source ',Cdotwrite(13,*}' '

Read(11,*)nlpb,nlpbp,nlpr

if(nlpb.gt.1}thendo k=2,nlpbRead(ll,*)Lb(k)

enddoendif

if(nlpbp.gt.1)thendo k=2,nlpbpRead(ll,*)Lbp(k)

enddo

endif

if(nlpr.gt.l}then

do k=2,nlprRead(11,*)Lr(k)

enddoendif

c The following acounts for Wfeed losses to other than turbinec building and reactor building losses.

Lxtra=O.

if(nlpb.gt.l)thendo k=2,nlpbLxtra=Lxtra+Lb(k)

enddo

endif

if(nlpbp.gt.l)then

do k=2,nlpbpLxtra=Lxtra+Lbp(k)

enddoendif

if(nlpr.gt.l)thendo k=2,nlpr

Lxtra=Lxtra+Lr(k)enddoendif

Call EVAP

Lb(1)=Wfeed-Edot(1)-LxtraLbp(1)=Edot(1)Lr(1)=Edot(2)

do k=l,nlpb

write(13,*)' loop leak rate no.',k, ' at power',Lb(k),' lbm/day+,

enddo

do k=l,nlpbpwrite(13,*)' pool leak rate no.',k,' at power',Lbp(k),' ibm/day

+,

enddo

C_24

do k=l,nlpr

write(13,*)' Ip+pl leak rate no.',k,' refueling°,Lr(k), '+ ibm/day'enddo

c tb=buildup period, daysc tr-refueling period, daysc Ncm=no. of cycles

c dtb=time step during buildup, daysc dtr-time step during refueling, daysc Cdot=H3 source rate to coolant, Ci/dayc Mo=coolant water inventory, pounds.c Mr=refuel pool inventory, poundsc decay lamda in 1/day for H3. Idk=l.53912E-04

Idk=l.53912E-4

c leak rate sumstlrb=0.tlrr=0.

tlrbp=0.

do ik=l,nlpbtlrb=tlrb+Lb(ik)

enddo

do ik=l,nlpbptlrbp=tlrbp+Lbp(Ik)

enddo

do Ik=l,nlprtlrr=tlrr+Lr(lk)

enddo

c Start time marching sequence.

nc=0

ncalcb=INT(tb/dtb)ncalcr=INT(tr/dtr)blr=Idk+tlrb/Mo

rlr=idk+tlrr/(Mo+Mr)blrp=Idk+tlrbp/Mr

DO jt=l,Ncm

do k=l,nlpbdCb(k)=Cb(k)

enddo

do k=l,nlpbpdCbp(k)=Cbp(k)

enddo

do k=l,nlprdCr(k)=Cr(k)

enddo

nc=nc+l

t=0.

write(12,*)' 'write(12,*)' 'write(12,*)'CYCLE NUMBER (',NC,')'WRITE(12,*)' 'write(13,*)' 'write(13,*)' 'write(13,*)'CYCLE NUMBER (',NC,') ************************

write(12,*)' Days Coolant H3, Ci Pool H3, ci'

_.25

Do mb=l,ncalcb Ot=t+dtb

q=Cdot/blrp=(-blr*CM3+Cdot)/(-blr)aa-blr

Lkb-q*dtb+p*(1.-exp(-aa*dtb))/aaLkbp=(CH3p/blrp)*(1.-exp(-blrp*dtb))write(13,*)' '

write(13,*)' cycle day= ',t,' (at power)'c loop leak

do J-l,nlpbCb(J}-Cb(j}+Lkb*Lb(j}/Mowrite(13,*)' Total loop leak from path',j,' ",Cb(J),' CA'

enddo

c pool leak

do j=l,nlpbpCbp(j)=Cbp(j)+Lkbp*Lbp(j)/Mr

' ' Cbp(j),' Ci'write(13,*)' Total pool leak from path',j, ,enddo

sumdk=sumdk+Lkb*Idk+Lkbp*idk

CH3=(-Cdot+(-blr*CH3+Cdot)*exp(-blr*dtb))/(-blr)CH3p=CH3p*exp(-blrp*dtb)Write(12,*)t,CH3,CH3p, ° at power'

Enddo

do k=l,nlpbsumlk=sumlk+Cb(k)

enddo

do k=l,nlpbpsumlk=sumlk+Cbp(k)

enddo

CiTOT=CH3+CH3pCH3=Mo*CiTOT/(Mo+Mr)CH3p=Mr*CiTOT/(Mo+Mr)

Do mp=l,ncalcrt=t+dtr

Lkr=((CH3+CH3p)/rlr)*(l.-exp(-rlr*dtr))write(13,*)' 'write(13,*)' cycle day= ',t,' (refueling)'do j=l,nlprCr(j)=Cr(j)+Lkr*Lr(j)/(Mo+Mr)

' ' ' Ci'' Cr(j),write(13,*) Total pl+lp leak from path',j, ,enddo

sumdk=sumdk+Lkr*ldk

CH3=CH3*exp(-rlr*dtr)CH3p=CH3p*exp(-rlr*dtr)

Write(12,*)t,CH3,CH3p,' refueling'Enddo

do k=l,nlprsumlk=sumlk+Cr(k)

enddo

source=float(nc)*Cdot*tb

ii{

netdif=source-CH3-CH3p-sumlk-sumdkwrite(14,*)' 'write(14,*)' cycle number ',ncwrite(14,*}' source total ',sourcewrite(14,*}' coolant inventory ',CH3

write(14,*)' refuel pool inv_y ',CH3pwrlte(14,*)' Cl leaked ',sumlk

write(14,*}' Ci decayed ',sumdkwrite(14,*}' net difference ",netdifsumlk-0.

write(13,*}' 'write(13,*)'Cycle ',jr,' Incremental Leak Values:'write(13,*)" 'do k=l,nlpbdCb(k}-Cb(k)-dCb(k)

' ' dCb(k} ' Ci'write(13,*)' Cycle loop leak from path',k, ,enddo

do k=l,nlpbpdCbp(k)=Cbp(k)-dCbp(k)

' ' dCbp(k) ' Ci'write(13,*}' Cycle pool leak from path',k, ,enddo

do k=l,nlprdCr(k}=Cr(k)-dCr(k)write(13,*)' Cycle pl+Ip leak from path',k,' ',dCr(k),' Ci'

enddo

ENDDO

C The final output is a compilation of Ci leaked to theC environment or directed out of the coolant-pool water inventory.

C The final cycle losses to the turbine building during power isc dCb(1), to the reactor building during power is dCbp(1), and toc the reactor building during refueling is dCr(1). All other dC'sc are "other" streams. The f_rmer terms enter the buildings' hvac.

RBconcap=dCbp(1)/(tb_24.*60.)/l.E-O6/RBacfm/28316.85RBconcrf=dCr(1)/(tr*1440.}/l.E-06/RBacfm/28316.85TBconcap=dCb(1)/(tb*1440.)/l.E-O6/TBacfm/28316.85tt=tb+tr

write(15,*)' 'write(15,*)' 'write(15,*)' TRITIUM DISPOSITION OUTSIDE REACTOR'

write(15,*)' 'write(15,*)' 'write(15,*)'The vaAues printed below can be used to estimate offsi

+re doses (from building'

write(15,*}'losses) and occupational doses (from building airborne+ concentrati_.ns}. Doses'

write(15,*)'must be computed for tritium as part of a water molecu+le (TOH or T20), not as'

write(15,*)'T2 or HT. If tritium is the only contributor to occup+ational dose, then a'

write(15,*)'limit of 2E-06 microCi/cc can be tolerated before resp+iratory protection is'write(15,*)'required.'write(15,*)' 'write(15,*)' '

write(15,*)' The following Ci and concentrations are for the tw+o periods in the'

write(15,*)tt,' day production cycle number',Ncm,' with '

C- 27

write(15,*)tb, ° days at power and',tr,' days refueling. 'write(15,*)" 'write(15,*)' '

write(15,*)' Reactor Building:'write(15,*}' '

write(15,*)' Total evap. loss during power = ',dCbp(1},' Ci'write(15,*)' Air H3 conc during power - ',RBconcap,' microCi

+/co'write(15,*)' '

write(15,*)' Total evap. loss refueling - ',dCr(1),' Ci'write(15,*}' Air H3 conc refueling - ',RBconcrf,' microCi/cc

+,

write(15,*)' '

write(15,*}' Turbine Building:'write(15,*)' •

write(15,*}' Total steam loss during power = ',dCb(1),' Ci'write(15,*}' Air H3 conc during power = ',TBconcap,' microCi

+/cc'write(15,*)' 'write(15,*)' Other Losses:'write(15,*}' 'if(Lxtra.eq.O.)thenwrite(15,*)' No other losses'

endif

if(nlpb.gt.l)thendo k=2,nlpb

' =' dCb(k) ' Ci'write(15,*)' Loop leak',k, , ,enddo

endif

if(nlpbp.gt.l)thendo k=2,nlpbp

' =' dCbp(k) ' Ci'write(15,*)' Pool leak',k, , ,enddo

endif

if(nlpr.gt.1}thendo k=2,nlpr

write(15,*)' Refueling leak',k,' =',dCr(k),' Ci'enddo

endif

creturnEND

C ************************************************************************

Subroutine INGRATE(rtau,ts,rint)

Tl=rtau*86400.t =.0100503,T1

i_(ts.le.t )rint=.01*ts

if(ts gt t_)thenal=.01*t

a2=(ts-t_)+Tl*(exp(-ts/Tl)-exp(-t_/Tl))rint=al+a2

endif

RETURNEND

C ************************************************************************

C-28

SUBROUTINE GETARG

REAL lambdaT, len,N_Anrg,N_Arate,mwllalo2,1t,mgbol,+mgeol,Li6depl,d(0_11),rcm(0_11},A(lO),N A

COMMON/BLK1/Nt,A,d,len,plen,N_A,ntgr,TGASR,gammaen,enrichli,+FPcycle,EOcycle,pT2,TpelletCOMMON/BLK5/tfluid,tcladfCOMMON/BLK6/YSfluid,YS10OF,PfluldCOMMON/BLKIO/GVR,GVRo

open(Unlt=4,File-'t2p2-3.out',status='unknown'}

c Data d/.5893,.6096,.9906,1.0084,1.049,1.0592,1.0744,1.2268,4,0./c nt=7

do l=0,Nt

rcm(1)=d(1)/2.enddo

teol=S6400.*FPcycle

rcladod=rcm(7)rcladid=rcm(6)ralumid=rcm(5)rgetod=rcm(4)rgetid=rcm(3)rpelod=rcm(2)rpelid=rcm(1)rlinrid=rcm(0)

pi=3.14159

fen=381.

plen=25.4lambdaT=l.792E-9

N_Anrg=4.8*l.6E-13N Arate=N A*len

c FPcycle=284.c EOCycle=378.

TGr=460.+600.

if(ntgr.eq.2)TGr=TGASRVcladid=pi*ralumid**2.*(len+plen)ccor=ii.94/12.

Vpellet=pi*(rpelod**2.-rpelid**2.)*len*ccorVgetter=pi*(rgetod**2.-rgetid**2.)*len*ccor

Vget Ni=pi*((rgetod-.00127)**2.-(rgetid+.00127)**2.)*len*ccorVliner=pi*(rpelid**2 -rlinrid**2.)*len*ccor

Vspring= 3.07Vnet=Vcladid-(Vliner+Vgetter+Vpellet+Vspring)

VVR=Vnet/VpelletTotrxns=N Arate*teol

T2molec=Totrxns/2.Heatoms=Totrxns

totmoles=(T2molec+Heatoms)/6.023E+23GVR=totmoles*22414./VpelletGVRo=exp(6.44-Tpellet*4.29E-03)sat2=9720.* 6.0321

c

if(ntgr.eq.l} goto 10write(4,*}' 'write(4,*)' 'write(4,*}' ********GVR/PRESSURE/STRESS CALCULATIONS********'write(4,*}' '

wrlte(4,*}'Thls output focuses on the end of the target irradiatlo+n period (eol)'write(4,*}'in the reactor. At this point the Anterior pin pressur

+e is at maximum'

write(4,*)'while still at power with the coolant at maximum temper+ature. The'

write(4,*}'coolant is then cooled to 100F. Symbol definitions:'wrlte(4,*)' 'write(4,*)' VVR - void volume/pellet volume'write(4,*}' 90%YS = 90% yield stress'write(4,*)' TH/Zr ratio = atoms of H+T/atoms Zr'write(4,*)' 'write(4,*)' '

e , t , #write(4,*)' Vcladid ',Vcladid, Vpellet ,Vpellet, cm* 3write(4,*)' Vgetter ',Vgetter,' Vget Ni ',Vget Niwrite(4,*)' Vliner ',Vliner ,' Vsprlng ,Vspr_ngwrite(4,*)' Vnet ',Vnet ,' VVR ',vvrwrite(4,*)' GVR ',GVRwrite(4,*)' GVRo ',GVRo

I0 continue

rhliaio2=2.57

mwlialo2=enrichli*6.02+(l.-enrichli)*7.02+26.89+32.gmlialo2=Vpellet*rhlialo2/mwlialo2atomsLi6=gmlialo2*6.023E+23*enrichli

Li6depl=100.*Heatoms/atomsLi6ccf=lO0.*FPcycle/EOcycleif(ntgr.eq.l)go to iiwrite(4,*)' '

write(4,*)' Li6 depletion % ',Li6deplwrite(4,*)' '

write(4,*)' Reactor Cycle Capacity Factor ,ccf,ii continue

totHe=Totrxns

totT2ndk=TotHe/2.It=l_,bdaT*teol

totT2wdk=totT2ndk*(l.-exp(-it))/itgmoleT2=totT2wdk/6.023e+23gmoleH2=0.1*gmoleT2

gmoleZr=Vget_Ni*6.4/91.22TH Zrrat=2.*(gmoleT2+gmoleH2)/gmoleZrm

CiT2=sat2*gmoleT2

c Fraction of T2 in gas space, frgasT2pT2=pT2/I.01325E+5

gasT2=pT2*Vnet/82.06/(TGr/l.8)totT2=totT2ndk/6.023E+23frgasT2=gasT2/totT2

if(ntgr.eq.1)go to 20write(4,*)' '

write(4,*)' T2 molecules, no decay ',totT2ndkwrite(4,*)' T2 molecules, w/ decay ',totT2wdk

' ' TH/Zr ratio ',TH Zrratwrite(4,*)' T2 curies ,CiT2, m

write(4,*)' Fraction of T2 in gas ',frgasT2

C'-- 30

20 continue

Aclad=7.92*gammaenAliner=6.4*gammaen

Agetter-gammaen*((Vgetter-Vget_ni}*8.9+Vget Ni*6o4)/VgetterAlialo2=gammaen*2.57 + N_Arate*NAnrg/Vpelletif(ntgr.eq.1)go to 30write(4,*}' 'write(4,*}' heat generation watts/cm**3: 'write(4,*)" cladding ',Aclad

write(4,*}' getter ',Agetterwrite(4,*)' liner ',Alinerwrite(4,*)' pellet ',Alialo2

30 continue

A(1)=AgetterA(2}-Alialo2A(3}=O.A(4}=AgetterA(5)=0.A(6)=AcladA(7)=A(6)

pressI=(GVR*2./3./VVR+I.)*TGr/492.mgbol=Vnet/22414.mgeol=mgbol+totHe/6.023E+23press2=mgeol*82.06*(TGr/l.8)/Vnetpress3=14.7*mgeol*82.06*560./l.8/Vnet

pressl=14.7*presslpress2=14.7*press2rd=(rcm(7)+rcm(5))/2.

CALL STRESS

ysg_l=.9*YSfluid*1000.ys9_2=.9*YSIOOF*I000.presnetl=-Pfluid+press2-14.7Pfi00=12.9159

presnet2=press3-PflO0-14.7hoopstl=presnetl*rd/(rcm(7)-rcm(5))Axialstl=hoopstl/2.hoopst2=presnet2*rd/(rcm(7)-rcm(5))Axialst2=hoopst2/2.

C Calculation of puff releases of gas phase tritium with sudden failureC of pin cladding, press4 ks at 75F.

press4=press3*535./560.

gasCi=CiT2*frgasT2puffl=(press2-Pfluid-14.7)*gasCi/press2puff2=(press3-Pfl00-14.7)*gasCi/press3puff3=(press4-14.7)*gasCi/press4

if(ntgr.eq.l)go to 40write(4,*)' '

write(4,*)' internal pressure at power',press2,' psi'

write(4,*)' internal pressure at 100 F',press3,' psi'write(4,*)' '

write(4,*)' hoop stress at power',hoopstl,' psi'write(4,*)' axial stress at power',axialstl,' psi'write(4,*)' '

' ' psi'write(4,*)' hoop stress at 100 F ,hoopst2,writ (4,*)' axial stress at 100 F',axialst2,' psi'

(- 31

write(4,*)' '

write(4,*}' 90%YS@pwr ',ys9 1,' 90%¥S100F ',y89_2,' psi'' netDP100F ',presnet2,write(4,*}" netDP@pwr ',presnetl, ' psi'

write(4,*}' '

write(4,*} ° Puff T2 Ci released from sudden eol pin failures'write(4,*}' ',puff1,' at power to coolant'

' ' ' @ 100F in pool'write(4,*) ,puff2,

write(4,*}' ',puff3,' @ 75F in air at 1 atm'40 continue

returnend

C ************************************************************************

Subroutine EVAP

C EVAP calculates pool water loss and water concentration of the ABWRC reactor building airspace.

Real Edot(2)COMMON/blk3/edot,rbacfm

Ci=3.19E-5

Cp=l.58E-4

Ak=3531.47,3.c Ak An cfm

c Ci = inlet air water conc., #moles/ft3. (50% rh at 70 F)c Cp= pool surface vapor conc (100% rh at 100 F)c Ak = mt coeff * pool area, cfmc A=600 m2

c k=30OOcm/hr (Fig.15-5, Lyman, W.J., et al. 1982. "Handbook of Chemical

c Property Estimation Methods." McGraw-Hill, NY.)c Co= reactor outlet vapor conc.,Ibmole/ft**3c Edot= evaporation rate, ibm/day

do j=2,1,-1acfm=rbacfm

Co=(Ci*acfm+Ak*Cp)/(acfm+Ak)

c Co_=Ak*Cp*lS./(acfm+Ak)/28316.85c Co_=pool added vapor conc, ibm/cc

Edot(j)=Ak*(Cp-Co)*18.*1440.Co =Edot(j)/1440./acfm/28316.85

c write(2,*)acfm,Co,Co_,Edot,AkAk=Ak/2.

enddo

return

end

C ******************************************************************

subroutine FAIL

c Program FAIL calculates the target pin-to-coolant source term asc fraction of tritium produced.

REAL Loss(20,20),Tbp(21},F(20),Fi(20),FR(20,20),npinsreal A(lO),d(O:11),n aCOMMON/BLK4/FailT2fr?rrctrc,tb

COMMON/BLK1/Nt,A,d,len,plen,N_A,ntgr,TGASR,gammaen,enrichli_+FPcycle,EOcycle,pT2,Tpellet

d 32

COMMON/BLKI0/GVR,GVRo

c nsdpc - number of undesired shutdowns per cyclec tap = days at full power in target cyclec npp - number of periods of power/cycle - nsdpc+l

nsdpc=3*nint(rrctrc)npp=nsdpc+nint(rrctrc)tap-FPcycleif(rrctrc.lt.l.01)thennsdpc=2npp=nsdpc+l

endif

npp=nsdpc+lnpins=3488.

c dos = day of pellet saturation with tritiumdos=tap*GVRo/GVR

c nps = period of power where saturation occursnps=npp

tpp=Tap/float(npp)c tpp = days of power period between shutdowns

Tbp(1)=O.do j=2,npp+lTbp(j)=Tbp(j-l)+tppif(Tbp(j).gt.dos.and.Tbp(j-l).le.dos)nps=j-i

enddo

c Define X, the fraction of period nps that is unsaturated.split=dos-float(nps-l)*tppX=split/tpp

c Define Loss(J,K) = fraction of tritium produced in K period due toc failure in J period.

DO J=l,npp

Do K=l,nppif(K. le.J)Loss(J,K)=0.if(K.gt.J)Loss(J,K)=0.03

if(K.gt.J.and.K.eq.nps)Loss(J,K)=.03*X+l.-Xif(K.gt.J.and.K.gt.nps)Loss(J,K)=l.

Enddo

ENDDO

c Weibull Statistics for failure frequency:

c Define F(J) as cumulative probability of failure over interval J.

c (see Mitchell, R.A..1967. Introduction to Weibull Analysis. PWA3001.c Pratt & Whitney Aircraft, E. Hartford, CT.)

c Weibull parametrs beta, eta, to.beta=l.045

c beta from Trans ANS 18, p125 (1974).to=Tbp(2)

eta=(3.*tb-to)/exp(-9.21029/beta)

c above eqn assumes failure rate of 1/10000/3 cycles(same as fuel pin).

F(1)=0.

C- 33

Fi(1)-O.do j=2,nppFi(J)=l.-exp(-((Tbp(J+l)-to)/eta)**beta)F(J}=Fi(J)-Fi(j-I)

enddo

FT=0.

do j=l,nppFT=FT+F(J)

enddo

T21oss=O.

Do j=l,nppdo k=l,nppFR(j,k}=Loss(j, k}*F(j )T21oss=T21oss+FR(j,k)

enddoEnddo

T21oss=T21oss/float(npp)FailT2fr=T21oss

c T21oss = fraction of tritium lost/cycle due to pin failure.

returnend

***********************************************************************

SUBROUTINE STRESS

COMMON/BLK5/tfluid,tcladfCOMMON/BLK6/YSfluid,YSIOOF,Pfluid

c Vapor pressure of water at temperature.

AAP=3.2437814BP=5.868263E-3CP=1.17023793E-08DP=2.1878462E-3a0=96.11825al=-3.02965E-2

tf=tfluid

tk=(TF-32.)/l.8 + 273.16

xp=647.27-tk

pvatm=218.167*lO.**(-(xp/tk)*(aap+bp*xp+cp*xp**3.)/(l.+dp*xp))

pfluid=14.7*pvatm

ysfluid=a0+al*tcladfyslOOF=91.44

returnend

C" 34

APPENDIX DRADIOLOGICAL SAFETY REQUIREMENTS AND CRITERIA

I. GENERAL

This section presents the safety standards, regulations and criteria agaimt which SMP will be comparedin order to demomtrate safety adequacy. In summary, the design, construction, commissioning andoperation of SMP will be carried out in accordance with the requirements of the THORP Division SiteLicense Regulations (SLRs). In addition the following specific regulations and criteria are to be appliedto SMP'

a) Radiological Protection Regulations (RPRs) - THORP Division; these will apply to normaloperational exposure.

b) Environmental Protection Regulations (EPRs) - THORP Division; these will apply to environmentaldischarges.

c) Radiological accident risk criteria for Sellafield reprocessing divisions and Drigg; these will applyto accidents.

The standards, regulations and criteria applicable to normal operations are summarized in Sectiom 2, 3and 4. The criteria for accidents are summarized in Section 5.

II. OCCUPATIONAL RADIATION EXPOSURE CONTROL

The operational radiation exposure in SMP will t_e in accordance with the requirements of the THORPDivision RPRs. The main requirements are summarized below:

* All the operations which lead to whole body exposure must be shown to be ALARP.

• The average radiation exposure, of the group of workers associated with the plant, should notexceed 5 mSv whole body dose (the sum of effective dose equivalent (EDE), from external radiationand committed effective dose equivalent (CEDE), from ingestion of radioactive material) per

calendar year, unless a special justification on ALARP grounds is authorized by the appropriateDivisional Director.

• The maximum whole body dose (EDE plus CEDE) for any individual worker should not exceed

15 mSv in any one year.

• The maximum extremity dose for any individual should not exceed 300 mSv per year.

III. ROUTINE AERIAL EFFLUENT DISCHARGES

The routine aerial effluent discharges from SMP will be in accordance with the requirements of theTHORP Division EPRs. This requires that all radioactive aerial discharges must comply with theconditions of the Discharge Authorizations. Specific requirements include:

• All plant will be designed and operated using Best Practicable Means to ensure that doses toindividual and collective doses are ALARP.

Appendix D- 1

• The quantities of radionuclides discharged to atmosphere from specified discharge points in anyperiod of 24 hours will be limited.

• For the identified discharge points the quantities of radionuclides discharged to atmosphere in anyquarter above specified levels are notifiable.

• The quantities of radionuclides discharged to atmosphere in any calendar year are limited.

• Sampling and measurement arrangements for discharge accountancy are specified.

• Formal records of discharges are maintained, then summarized, and forwarded to the AuthorizingDepartments within specified time periods,

At this stage of the project, SMP has not been allocated a discharge target. However, SMP will need

to be designed to take account of these requirements and will ultimately receive a discharge allocationin line with the requirements: this allocation constituting a percentage of the THORP division dischargeallocation.

IV. ROUTINE LIQUID EFFLUENT DISCHARGES

The routine liquid effluent discharges from SMP will be in accordance with the requirements of theTHORP Division EPRs. This requires that all discharges to sea of low active liquid waste from theSellafield site must comply with the conditions of the Discharge Authorizations. Specific requirementsare that:

• All plant will be designed and operated using Best Practicable Means to ensure that doses toindividual and collective doses are ALARP.

• The quantities of radionuclides discharged to sea in any period of 24 hours are limited.

• The quantities of radionuclides discharged to sea in any quarter are limited.

• The quantities of radionuclides discharged to sea in any calendar year are limited.

• Sampling and measurement arrangements for discharge accountancy are specified.

• Formal records of discharges are maintained, then summarized, and forwarded to the AuthorizingDepartments within agreed time periods.

At this stage of the project, SMP has not been allocated a liquid effluent discharge target. However,SMP will ultimately receive a discharge allocation in line with the above requirements; this allocation willcontribute to the overall THORP division discharge allocation.

V. ACCIDENT CONDITIONS

The risk criteria against which accidents associated with SMP will be compared are summarizedbelow. In terms of these criteria SMP is considered as 'one plant'.

Appendix D -2

V.A Accidental Risk Criteria for the Workforce (Internally Initiated Events)

a) The summed frequency of accidents (including criticalities) in a building, which could give dosesto any member of the workforce working in that building, from direct radiation, inhalation andingestion should be less than the values given below:

Effective Dose (mSv) Summed Frequency(per year)

.... 50"- 10()0 ' 10 .3

> I000 10.5L.....

b) The summed frequency of Building Evacuations due to abnormal airborne contamination (exceeding100 DAC) and high gamma dose rates (exceeding 0.2 mSv hr j) in large parts of an operating areaof a plant should be less than 10.2 y_.

c) The summed frequency of unplanned criticality events should be less than 104 yJ.

Interpretation, For the purpose of these criteria, an internally initiated event is defined as an event withan initiator on the plant or any other plant on the site where the effect on the plant being assessed is via

process or service routes. Consequently, events initiated by local road and rail vehicles, cranes andmobile platforms should be included. External events can be naturally occurring (e.g., earthquakes),man-made but outside the Company's control (e.g., aircraft crashes) and catastrophic failure of otherplant on the site (e.g., a turbine failure) where the initiator is not linked to the plant being assessed viaa process or service route.

These criteria apply to buildings, as opposed to the 'plant' concepts used fc:" accidental releases affectingthe public. However, a building should include subsidiary and ancillary buildings which would beregarded as part of a main process building as far as the process operators would be concerned. (Thiswill usually coincide with the scope section of a safety case.)

The criterion in V.A (b) above, applies only within buildings. It is not applicable to restricted accessareas, where higher airborne activity or radiation levels are to be expected and appropriate precautionstaken. The other criteria apply to all areas.

10_sfissions should be used for the reference criticality event in estimating consequences - V.A(a), unlessanother more appropriate value is agreed with the Site Nuclear Safety Officer.

Doses from contaminated wounds are not included in these criteria: qualitative assessment will suffice.

V.B Criteria for Accidental Aerial Discharges

a) The time averaged critical group dose from accidental aerial discharges from a plant should notexceed 4 #Sv per year.

Appendix D-3

b) The summed frequency of accidents on a plant, which would give doses to a member of the criticalgroup, should be less than the values given below:

--

Effective Dose (mSv) Summed Frequency(per year)

0.01 - 1 10"z

1 - 10 10.3

10- 100 104

100- 1000 10.5

> 1000 10.6......

c) The summed frequency of very large accidents on a plant, i.e., those with the potential to give arelease to the environment with societal consequences equivalent to:

- 100 deaths, e.g., a release 10,000 TBq of Iodine 131, 5 TBq Pu.or

- Land contamination effects equivalent to that from 200 TBq of Caesium 137.

should be less than 10 .6 pa.

Interpretation. For the purpose of these criteria, an internally initiated event is definec as an event with

an initiator on the plant or any other plant on the site where the effect on the plant being assessed is viaprocess or service routes. Consequently, events initiated by local road and rail vehicles, cranes andmobile platforms should be included. External events can be naturally occurring (eg earthquakes), man-made but outside the Company's control (eg aircraft crashes) and catastrophic failure of other plant on

the site (eg a turbine failure) where the initiator is not linked to the plant being assessed via a process orservice route.

In setting the criterion in V.B(a) above, allowance has already been made for the variations which occurin wind direction (wind direction is not relevant to the other criteria - Paras (b) and (c).

Events having consequences less than 10 _Sv (all pathways) should be ignored. For effective doses upto 10 mSv, all pathways dose should be used. For effective doses greater than 10 mSv, inhalation anddirect radiation doses only should be used - Paras (a) and (b).

V.C Criteria for Accidental Liquid Effluent Discharges

a) The time averaged critical group dose from accidental marine discharges from a plant should notexceed 0.4 pSv per year.

Appendix i") .4

b) The summed frequency of accidents on a plant, which could give doses to a member of the publicoutside the site arising from discharges to sea or waterways, directly or via engineered routes,should be less than the values given in the following table:

Effective Dose (mSV) Summed Frequency(per year)

.... 1o"'-loo 1or

> I00 10"4

Interpretation. For the purpose of these criteria, an internally initiated event is defined as an event withan initiator on the plant or any other plant on the site where the effect on the plant being assessed is via

process or service route. Consequently, events initiated by local road and rail vehicles, cranes and mobileplatforms should be included. External events can be naturally occurring (eg earthquakes), man-madebut outside the Company's control (eg aircraft crashes) and catastrophic failure of other plant on the site(eg a turbine failure) where the initiator is not linked to the plant being assessed via a process or serviceroute.

The criteria apply to all accidental liquid discharges, not just those via the sea lines. However, eventshaving consequences less than I0 _tSv should be ignored,

Assessment of consequences and frequencies should include consideration of clean-up plants outside thesubject plant, their reliability, and the possibility of detection and associated diversion of liquors outsidethe plant on Site where appropriate.

V.D Accident Risk Criteria for Externally Initiated Events

Seismic Events

a) Plant design shall be examined without seismic provisions to determine the expected dose to amember of the public from inhalation and direct radiation following a design basis seismic event(such an event shall be defined as having an annual probability of exceedence of 104).

- Where the dose is greater than 5 roSy, then the plant will be seismically qualified to show that

the expected dose to a member of the public off site is less than 5 roSy following the designbasis event.

- Where the expected dose to a member of the public off site is greater than I mSv but less than5 roSy, then the plant should be assessed to show that it will withstand an event with annual

probability of exceedence of 103.

- Where the dose is less than I mSv, no special seismic provisions are required beyond the

general requirement to satisfy ALARP.

b) Seismically qualified plants designed to Paragraph (a) should be examined to show there are nocliff-edge effects on release for events beyond the design events specified. This can be

demonstrated by considering a more severe event. If this cannot be demonstrated, appropriateaction should be taken where this is shown to be reasonably practicable.

Appendix'D-5

c) Where it is considered feasible that acute health effects (doses > I Sv) to workers may occurfollowing a seismic event, then the plant should be assessed to show that following a seismic eventwith annual probability of exceedence of I03, personnel escape routes from areas with highoccupancy are likely to be available.

Extreme Wind

a) It should be shown by deterministic argumentsthat containment barriers remain and any safetysystems which are required during or after the event can survive an extreme wind with an annualprobability of exceedence of 104, for those plants with potential to give a dose to a member of thepublic off site more than 5 mSv in the event of such a wind.

Notes:

- This may entail considerationof containment structures, operatingarea structures(eg buildingshell), ventilation, cooling services etc.

- This may be achieved by design and assuring that vulnerable inventories or plant areprotected, or can be protected given advanced notice of the event, from damage.

b) Plant with radioactive material directly exposed to the wind (such as, ttdoor ponds), should beexamined against extreme wind with an annual probability of exceedence of 10'*yt and it be shownthat the consequences so far as is reasonably practicable to a member of public off site are less than5 mSv from inhalation and direct radiation.

Precipitation

a) The direct effects of rainfall, local accumulation of rain water and rain falling on or around abuilding or site should be examined against site-specific data, taking account of interactions withabnormal tidal effects as appropriate. It should be shown that radioactive material is protected fromrainfall by a structure and is out of reach of flood water, or that facilities exist to containcontaminated run-off/overflow.

b) The impact of snow and its accumulation should be addressed deterministically, taking account ofcounter-measures where appropriate.

Temperatures and Drought

a) Extreme temperatures and drought do not occur unexpectedly. Where special provisions forpredicted extremes are not incorporated in design, it should be shown that contingency arrangementsto assure safety in extremes occurring within periods of abnormal temperatures and drought can beintroduced in a timely manner.

Aircraft Crashes

a) Where it can be demonstrated that the frequency of aircraft crashes with the potential to lead to arelease of radioactive material off site, taking account of pilot action and exclusion zones asappropriate, is not more than 10.7y.i per plant, then no further action to deal with aircraft crashesis required.

Appendix _I_-6

b) If the above criterion cannot be met then it should be shown that the conditions of paragraph (c)below are met by regarding an air crash as another ma" made external hazard.

Other Man Made External Events

a) Detailed consideration need only be given to the possibilities of off site explosion or gas cloudsaffecting the safety of plant if the site is within:

- 2km of a NIHHS notifiable site

The consultation planning approval distance of a NIHHS notifiable site

- 2km of a regularly used transport route for the movements of materials with the potential foraffecting the site

- Or 2km of a pipeline with a potential for affecting the site.

b) The effects of a catastrophic failure of the plant being assessed on other plants on the site (wherethe initiator is not linked to the plant via a service or process route) should meet the criteria in

Paragraph (c) or (d) below.

c) No action should be taken if it can be demonstrated that the frequency of any event affecting the

plant likely to lead to the release of radioactive material is less than I0 7 y.i, or the potentialexpected dose to a member of the public following the event is less than I mSv from a plant (takingaccount of the potential amount and type of external substance involved, the distance from the plant,

topography, wind direction and other relevant factors).

d) Where Paragraph (c) cannot be met it should be shown as far as reasonably practicable for eachexternal hazard that the best estimate consequences to a member of the public off site from a plantare less than the values in the table below:

Predicted frequency, f, per Effective Doseyear per event

per plant

10.7 < f < 10.6 I SvI0 6 < f < I0 5 100mSv10.5 < t < I0 4 10mSv

I0-4 < f ImSv

e) Where application of this proves difficult, the advice of the Director of Health Safety andEnvironmental Protection should be sought so that criteria can be developed to suit the needs of the

specific case.

Appendix D -7

III