Response to Portion of NRC Request for Additional Information … · 2012-11-21 · Request for...

GE Energy

David H. HindsManager. ESBWR

PO Box 780 M/C L60Wilmington, NC 28402-0780USA

T 910 675 6363F 910 362 6363david.hinds@gecom

MEN 06-441 Docket No. 52-010

November 20, 2006

U.S. Nuclear Regulatory CommissionDocument Control DeskWashington, D.C. 20555-0001

Subject: Response to Portion of NRC Request for Additional InformationLetter No. 40 Related to ESBWR Design Certification Application -ESBWR Probabilistic Risk Assessment - RAI Numbers 19.1-15, 19.2-12, 19.2-17, 19.2-26, 19.2-33, 19.2-35, 19.2-37, and 19.2-38

Enclosure 1 contains GE's response to the subject NRC RAIs transmitted via theReference 1 letter.

If you have any questions about the information provided here, please let me know.

Sincerely,

David H. HindsManager, ESBWR

General Electric Company

MFN 06-441 of 2

Reference:1. MFN 06-222, Letter from U.S. Nuclear Regulatory Commission to David Hinds,

Request for Additional Information Letter No. 40 Related to ESB WR DesignCertification Application, July 5, 2006

Enclosures:1. MFN 06-441 - Response to Portion of NRC Request for Additional Information

Letter No. 40 Related to ESBWR Design Certification Application - ESBWRProbabilistic Risk Assessment - RAI Numbers 19.1-15, 19.2-12, 19.2-17,19.2-26, 19.2-33, 19.2-35, 19.2-37, and 19.2-38

cc: AE Cubbage USNRC (with enclosures)GB StrambackGE/San Jose (with enclosures)eDRF 0000-0060-3203

Enclosure 1

MFN 06-441

Response to Portion of NRC Request for

Additional Information Letter No. 40

Related to ESBWR Design Certification Application

ESBWR Probabilistic Risk Assessment

RAI Numbers 19.1-15, 19.2-12, 19.2-17, 19.2-26, 19.2-33, 19.2-35,19.2-37 and 19.2-38

MFN 06-441 Page 1 of 14Enclosure 1

NRC RAI 19.1-15

Justify that the core loading and burnup assumed in developing the source term is representativeof the way the ESB WR plant will be operated.

GE Response

The core loading inventory assumed in developing the source term is bounding for enrichmentand exposure for GEl4 fuel. It assumes an end-of-cycle equilibrium inventory, with a coreaverage exposure of 36 GW/MTU, a maximum discharge exposure of 58 GWd/MTU and apower density of 5.75 MWt/bundle constant power. This is representative of the way theESBWR will be operated.

DCD Impact

No DCD changes will be made in response to this RAI.


NRC RAI 19.2-12

Vacuum breaker leakage/failure is assumed to lead to loss of pressure suppression in Section8.2.1 of the PRA. However, it appears that vacuum breaker leakage/failure could also lead toloss of PCCS. The leakage rate at which the pressure suppression function and the PCCSfunction are compromised could be different. Specify the maximum leak area that could existbefore: (a) the pressure suppression function is compromised, and (b) the PCCS function iscompromised.

GE Response

It is correct that vacuum breaker leakage/failure could lead to loss of PCCS. The PRA modelincludes two success criteria for the Drywell to Wetwell vacuum breakers.

* For the pressure suppression function, 2 of 3 vacuum breakers must remain closed todirect LOCA flow through the suppression pool and 1 of 3 vacuum breakers must open torelieve vacuum.

* For the PCCS function, 3 of 3 vacuum breakers must remain closed to allow PCCS tofunction.

The allowable leakage was discussed in RAI 6.2-12. A sensitivity study was described whichfound the peak drywell pressure of a feedwater line break accident would approach the designpressure of 45 psig at 72 hours after the pipe break if the leakage size were increased toAPK = 100 cm2.

Sensitivities were run using MAAP to determine allowable vacuum breaker leakage to preventcontainment overpressurization during the blowdown phase of a LOCA. Long-termpressurization would be controlled by the PCCS function. The following table presents theresults of these runs. The peak pressure in the early phase of the accident did not reach theultimate pressure capability of the containment of 1.204 MPa (174 psig) for any of the runs. Thegraph below shows the upper DW pressure during the first hour for the FW line break case with3 Vacuum Breakers (VB) failed open. These results support the existing success criteria for thepressure suppression function.

Peak Pressure in Early PhaseEvent 1 VB Open 2 VBs Open 3 VBs OpenMS Line Break 673468 Pa = 925305 Pa = 1114256 Pa =

97.7 psia 134.2 psia 161.6 psiaFW Line Break 705601 Pa = 974073 Pa = 1165824 Pa =

102.3 psia 141.3 psia 169.1 psia

MFN 06-441Enclosure 1

Page 3 of 14

Feedwater Line Break - 3 Vacuum Breakers Failed Open

I-- Upper DW Pressure

1400000

1200000

1000000

800000

600000

400000

a-

2a-

2 0 0 0 00

O.OOE+00 5.OOE+02 1.00E+03 1.50E+03 2.00E+03

Time, s

DCD Impact


2.50E+03 3.OOE+03 3.50E+03 4.00E+03


NRC RAI 19.2-17

Steps PC/G-3 and PC/G-6 of the Emergency Procedure and Severe Accident Guidelines(EP/SAG) contain explicit instructions to vent the drywell or suppression chamber, respectively,given certain hydrogen/oxygen concentrations. Describe the vent lines that would be used forthis purpose. Clarify whether vented releases pursuant to these instructions are included withinthe Level 2 PRA, and if not, why not.

GE Response

As discussed in NEDO-33201, Revision 1, Section 8.1, the analysis of the radiolytic oxygenconcentration in containment is based on the methodology of Appendix A to SRP 6.2.5 andRegulatory Guide 1.7. Some of the key assumptions are as follows:

Reactor power = 102% of rated power

Generation rate (02) = 0.25 molecules/100eV

Initial containment 02 concentration = 4%

Allowed containment 02 concentration = 5%

Stripping of drywell non-condensable gases to wetwell vapor space

Fuel clad-coolant reaction up to 100%

Iodine release up to 100%

Adequate gas mixing throughout containment

The analysis results show that the time required for the oxygen concentration to increase to thede-inerting value of 5% is significantly greater than 24 hours for a wide range of fuel clad-coolant interaction and iodine release assumptions up to and including 100%.

Therefore, the Level 2 PRA does not take credit for using the vent lines to prevent unacceptablehydrogen and oxygen concentrations. Venting for pressure relief is modeled as an operatoraction, i.e., no mechanical faults. The Level 2 Release Frequency is estimated to be 2 E-10 peryear for events where the operator vents the containment for pressure relief.

The vent lines are in the design phase and design details are not available at this time.

DCD Impact



NRC RAI 19.2-26

Provide a comprehensive discussion of the items/features controlling the lower drywell waterlevel for the period before RPV failure and for the period after RPV failure. This discussionshould address event timing, sensor activation, flow paths into the lower dry well (LDW),drainage and vent paths, gutters, curbs, boil-off replenishment, squib valves and actuationsystems, eutectic valves, and operator instructions when to use and when not to use containmentsprays.

GE Response

There are a limited number of flow paths into the lower drywell in addition to a LOCA. The flowpaths include the following:

The main flow path is between the upper and lower drywell. This path is restricted at thevessel supports into 8 areas, which are approximately 1.8 m2 each. This is shown inFigure 21.3.2-4 ofNEDC-33201P Rev. 1.

Twelve deluge lines connect the BiMAC in the lower drywell floor to the GDCS pools.The deluge lines, described in Section 6.2.1.1.10.2 of DCD Tier 2 Revision 1, areisolated by squib-type valves, which are actuated upon detection of lower drywell floortemperatures exceeding setpoint values. The deluge lines do not require the actuation ofthe squib-actuated valves on the injection lines of the GDCS piping to flood the lowerdrywell. The routing of these lines is not complete at this time.

The spillover vent lines connect the annular space above the vessel supports and belowthe floor between the upper drywell and the wetwell. The inlet to each line is closed by aspecialized swing disc which is designed to open due to buoyant force when the waterlevel in the lower drywell rises above the valve height. The discharge of the spillovervent lines will be near the bottom of the suppression pool.

* The Containment Inerting System includes an exhaust from the lower drywell. This ventis not modeled in the PRA.

The floor of the lower drywell includes two sumps with pumps to collect leakage. These sumpsare not modeled in the PRA. These sumps are located on the edge of the pedestal floor outside ofthe BiMAC.

The openings in the horizontal drywell surface for the vertical vents may include a curb. Thiscurb will ensure leakage is directed to the lower drywell where it can be measured. Due to ex-vessel explosion phenomena, this assumption in the PRA provides bounding results bymaximizing the amount of water that would collect in the LDW during severe accidentsequences.


The GDCS pool tanks include overflow drainage paths which direct overflow of the GDCS poolsinto the suppression pool. This would be required if a GDCS injection valve does not open or theRPV does not de-pressurize. The PCCS would continue to condense steam from the drywellwhich be returned to the GDCS pools.

The GDCS deluge actuation is accomplished using a grid of thermocouples in the lower drywellbasemat. The thermocouples sense the presence of molten fuel on the lower drywell floor. SeeDCD Tier 2 Revision 1 Section 6.3.2.7.2.

After actuation of the GDCS deluge lines due to RPV failure and relocation of the core to thelower drywell, the decay heat will boil the deluge water. The resulting steam will be condensedby the PCCS and drained to the GDCS pools. The deluge lines remain open once activatedresulting in the condensate returning to the lower drywell through the BiMAC.

There are no eutectic valves in the ESBWR design.

The ESBWR does not use the drywell spray in the EPGs. The ESBWR severe accidentguidelines include the following: "Drywell spray can be utilized if drywell water level is lessthan 0.7m or lower drywell floor thermocouples indicate the presence of core debris on thedrywell floor. If drywell water level is greater than 0.7m and lower drywell floor thermocouplesdo not indicate the presence of core debris on the drywell floor, drywell sprays should not beutilized". See DCD Tier 2 Chapter 18.

The evaluation of lower drywell water level for the Level 1 results is described in NEDO-33201Rev 1 Section 7.2.5. Liquid LOCAs provide the only means for depositing a large amount ofwater in the lower drywell prior to RPV failure. In other events, significant amounts of water arenot expected until after RPV failure and actuation of the deluge.

DCD Impact



NRC RAI 19.2-33

The ROAAM treatment of the direct containment heating (DCH) phenomenon concluded thatcontainment failure from a DCH event is physically unreasonable. This, in large part, ispredicated upon the assertion that all 8 DPVs, all 18 SR Vs, and two ICs will have to fail torender the depressurization system non-functional (page 19.3-8 of DCD). Provide technicaljustification and analyses for this assumption regarding number offailures that must occur.

GE Response

The ROAAM treatment of DCH concludes that catastrophic containment failure is physicallyunreasonable based on the analysis of physical phenomena. NEDO-33201 Revision 1, Section21.3 concludes that the ESBWR containment can withstand bounding DCH pressure loads andthat catastrophic containment failure due to DCH is physically unreasonable. This conclusion isbased upon a quantification of the fragility of the containment to DCH, as discussed in Section21.3. The following phenomena support this conclusion:

* The upper drywell atmosphere can vent into the wetwell through a large vent area and aneffective heat sink,

* The drywell head is (externally) immersed in water, and essentially isolated, from theupper drywell atmosphere,

* The containment steel liner is structurally backed by reinforced concrete, which cannotbe structurally challenged by DCH.

This assertion is not based on the low conditional probability of a high reactor pressure at thetime of vessel failure.

DCD Impact



NRC RAI 19.2-35

Provide the following design data: (a) flow area, elevation and form loss coefficient for thejunction between the separators and RPV downcomer, (b) flow area for the junction betweenRPV steam dome and downcomer, (c) clarification whether the entire annular space between theshield wall and the RPV is filled with insulation material, and the thickness and thermalconductivity of the materials, and (d) representative thickness, surface area (one side) andmaterial for the structures in the lower drywell, e.g., CRD service machine and platform.

GE Response

(a) Detailed information on the design of the steam separators is provided in response to RAI21.6-76.

(b) Area between the RPV dryer and RPV wall: Nominal RPV ID is 7.112 m. The OD of thesteam dryer seal skirt is 6.9195 m. The resulting area is 2.12 in2.

(c) Annular space between shield wall and the RPV outer wall is not filled with insulationmaterial as access to the RPV outer wall is required for ISI purposes. The design of theinsulation is not complete at this time. However, based on ABWR RPV design it will bemetal reflective type insulation, approximately 75 - 100 mm thick, located next to theshield wall. The average heat transfer coefficient of the ABWR RPV insulation is 0.907W/m2oC.

(d) The design of the under vessel equipment is not complete at this time. The MAAPParameter file is based on input from the ABWR parameter file. The assumed area for thelower drywell lumped heat sink is 509.7 m2. The assumed mass is 149,482 kg and thematerial is carbon steel.

DCD Impact



NRC RAI 19.2-37

GE's 12/29/2005 response to RAI 19.2-2 [originally 19.2.4-1] indicates that the DCD willinclude a combined license (COL) applicant commitment that: "The COL applicant referencingthe ESBWR certified design will develop and implement severe accident management guidance,along with the required procedures and training, using the framework provided in DCD Chapter18, Appendix A. "However, the referenced 'framework" is simply a general discussion regardingthe ESBWR version of the Emergency Procedures and Severe Accident Guidelines. It does notaddress training (as implied in the COL Applicant commitment) and fails to address otherimportant aspects of a licensee severe accident management program. Industry guidanceregarding the severe accident management closure process for operating reactors and keysevere accident management elements are provided in Section 5 of NEI 91-04, Revision 1,"Severe Accident Issue Closure Guidelines. "Additional guidance to BWR licensees is containedin the "B WR Owners Group "Accident Management Guidelines Overview Document. "Althoughdeveloped for operating reactors, this guidance can also be applied to advanced reactors. Thediscussion and commitment regarding the accident management program for ESBWR shouldinclude a broader description of the severe accident management program and its elements,and/or appropriate references to these guidance documents and how they will be utilized by aCOL applicant.

GE Response

GE will provide to the COL Holder, the technical basis for a severe accident managementprogram, including emergency procedure guidelines (EPGs), to ensure core damage preventionand mitigation, including meeting off-site dose limits. GE will translate the plant design basesinto operation limitations and responses, which can be developed into procedural guidelines andtraining by the COL Holder. GE will use the ESBWR PRA, and other relevant information toconfirm that the plant design is compatible with the EPGs and Severe Accident ManagementProgram.

GE will identify existing systems and equipment, which may be useful as part of the severeaccident management program. This includes both safety and non-safety equipment onsite aswell as offsite. The COL Holder will develop procedures that will identify those actions to betaken to prevent and mitigate the effects of severe accidents, including:

* Preventing core damage.

* Recovering from core damage without vessel failure.

* Maintaining containment integrity.

" Minimizing offsite radiation releases.


Following TMI, Severe Accident Management was identified as an unresolved issue. Anindustry initiative committed utilities to access current capabilities to respond to severe accidentconditions and implement appropriate improvements. NEI 91-04 Rev. 1 describes the closureprocess, to develop plant specific SAM guidelines (SAMGs), to interface SAM guidelines withthe emergency plan, to incorporate SAM into training, and to evaluate new SAM information.NEI 94-01 Figure, "Severe Accident Management Closure Process," shows the ESBWR COLholder commitment to NEI 91-04 Rev. 1.

COL Holder commitments include:

Section 5.2 of NEI 91-04 Rev. 1, which contain implementation guidance relative to the formalindustry position on severe accident management that is bounding to all utility members. TheSevere Accident Management closure process for a given COL Holder is recommended toconsist of the following steps:

" Evaluate industry-developed bases and Owners Group severe accident managementguidance (SAMG) along with the plant Individual Plant Examination (IPE), IPE ofExternal Events (IPEEE) and current capabilities, to develop severe accidentmanagement guidance for accidents found to be important in the plant as screened withthe criteria provided in Section 2.0 of NEI 91-04 rev. 1. Consider other generic and plant-specific information (e.g., NRC and industry studies, PSA results, etc.) as appropriate;

* Interface SAMG with the plant's emergency Plan:

" Incorporate severe accident material into appropriate training programs; and

" Establish a means to consider and possibly adopt new severe accident information fromlicensee self-assessments, applicable NRC generic communications, PRA studies, etc.

Section 5.3.1 of NEI 91-04 Rev. 1, Severe Accident Management Guidance/Strategies forImplementation, provides guidance for use by the ERO personnel in assessing plant damage,planning and prioritizing response actions, and implementing strategies that delineate actionsinside and outside the control room. Strategies and guidance will be interfaced with the ESBWREOPs and Emergency Plans.

Section 5.3.3 of NEI 91-04 Rev. 1, Computational Aids for Technical Support, states thatpersonnel should be provided computational aids, as appropriate, in estimating key plantparameters and plant response relative to accident management decisions.

Section 5.3.4 of NEI 91-04 Rev. 1, Information Needed to Respond to a Spectrum of SevereAccidents, provides guidance for awareness, and encouraged use, of instrumentation that isreasonable expected to be available fore assessing plant status. The availability and survivabilityof the information source and the ability of these sources to provide indication of sufficientaccuracy for the intended use will be considered. Alternative, indirect means for providingnecessary information will also be considered.


Section 5.3.5 of NEI 91-04 Rev. 1, Delineation of Decision-Making Responsibilities, ensuresresponsibilities for authorizing and implementing accident management strategies as part of theEmergency Plan.

Other COL Holder Commitments include:

Incorporate the BWROG Accident Management Guidelines Overview Document, Rev. 1,Section 6, four interrelated assessments as defined:

* The Control Parameter Assessment, which obtains and processes plant data.

* The Plant Status Assessment, which evaluates current plant conditions.

* The System Status Assessment, which evaluates the availability of systems needed toimplement EOPs and SAGs.

" The EPG/SAG Action Assessment, which prioritizes system restoration actions and

determines the appropriate timing of procedural actions.

The typical elements of the Severe Accident Management are as follows:

* Emergency Operating Procedures (EOPs)

Emergency Procedures that respond to emergencies and events that may degrade intoemergencies up until primary containment flooding is required.

* Severe Accident Guidelines (SAGs)

The SAGs define strategies applicable after primary containment flooding is requiredbased on the BWROG EPG/SAG Rev. 2.

Severe Accident Management Support Document

Provides a technical basis supporting severe accident management for the ESBWR. Theinformation is based on the severe accident analysis performed with the MAAP 4.0 code. Thisinformation relates to severe accident phenomena in the RPV (metal-water reaction, onset ofmelting, core relocation and RPV breach) and associated conditions in the containment(radionuclide and hydrogen distribution, and changes in pressure, temperature and suppressionpool water level). This severe accident management support document would include:

* Insights regarding the timing of key events for postulated severe accidents;

* Characteristic pressure and temperature profiles for a spectrum of postulated severeaccidents;

* Characteristics of suppression pool level response for a spectrum of postulated severeaccidents;

* Characteristics of core hydrogen generation and distribution in containment for aspectrum of postulated severe accidents;

* Insights regarding use of alternate injection sources such as AC-Independent Fire WaterAddition System to provide long term cooling


Functional Requirements Technical Support Guidelines

The Technical Support Guidelines (TSG) provide guidance to Emergency ResponseOrganization (ERO) personnel on supporting and optimizing accident management strategiesimplemented through plant Emergency Operating Procedures (EOPs) and Severe AccidentGuidelines (SAGs).

ESBWR Severe Accident Training

Training will be provided for the Emergency Response Organization (ERO) personnel with theirresponsibilities defined in the emergency Plan. Training to specific personnel that have severeaccident assessment and mitigation responsibilities will also be provided:

* Evaluators - Responsible for accessing plant symptoms in order to determine the plantdamage condition(s) of interest and potential strategies that may be utilized to mitigatean event.

* Decision Makers - in the emergency Response Organization (ERO) designated to accessand select the strategies to be implemented.

* Implementers - responsible for performing those steps necessary to accomplish theobjectives of the strategies (e.g., hands-on control of valves, breakers, controllers, andspecial equipment).

DCD Impact

In response to this RAI, DCD Chapter 18 will be revised to summarize and reflect thecommitments above in the next revision.


NRC RAI 19.2-38

Discuss whether the reactor water cleanup (RWCU) break outside containment described inPRA Section 9.1 bounds a temperature-induced IC tube failure or a water hammer inducedfailure of the IC tubes in terms offission product releases to the environment. Provide a briefassessment of these latter scenarios, including: (a) the probability of containment bypasssequence via an IC tube failure, and (b) the consequences of such a failure compared to theR WCU sequence analyzed in the PRA.

GE Response

The break outside containment in the IC, as an initiator, has been analyzed and did not contributeto core damage frequency.

Containment bypass due to an IC tube failure is not probable. A temperature-induced IC tubefailure requires that the level in the IC pool be lowered due to boiling which uncovers the IC heatexchanger(s). The IC heat exchanger is designed to withstand the design temperature andpressure of the RPV. The IC heat exchanger will not see higher pressures without multiplefailures of SRVs to control RPV pressure. Temperatures above the design temperature requirethat the core is first uncovered as steam exiting the core would be at saturation temperature.

Water hammer is not probable as the IC heat exchangers are normally pressurized due to theopen steam supply valves. Condensate fills the piping from the IC heat exchanger to thecondensate return valves. A loop seal between the condensate return valves and the RPV isdesigned to assure that steam continues to enter the IC heat exchanger preferentially through thesteam riser, irrespective of the water level inside the reactor, and does not move counter-currentback up the condensate-return line.

The consequences of an IC tube failure are bounded by the RWCU break outside containmentanalyzed in the PRA. The RWCU BOC sequence is an unisolated break outside of containmentin the shutdown cooling piping followed by no injection into the RPV. In this scenario, therelease begins at the onset of fuel damage and proceeds directly to the environment.

The release in the IC tube failure sequence would occur after fuel damage, as heat up of theuncovered IC heat exchanger is required. This sequence must be a Class III sequence (coredamage with the RPV at high pressure). This sequence also requires a failure to isolate the lines.

DCD Impact


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