Washington, JUN 2 7 2007 - Nuclear Regulatory Commission

Department of EnergyWashington, DC 20585

JUN 2 7 2007

Attn: Document Control DeskDirector, Spent Fuel Project OfficeOffice of Nuclear Material Safety and Safeguards.U.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001

This is in response to the request of additional information on Certificate of Compliance No.9315, Revision 5 for the Model ES-3100 Package, Docket No. 71-9315, TAC No. L24063. In aletter dated May 9, 2007, the Department of Energy received a request for additional informationconcerning the application for revision to the definition of allowable pyrophoric forms uraniumin the ES-3100 shipping package. These RAIs were discussed during a public meeting with theNuclear Regulatory Commission on May 10, 2007. The attached document, Y/LF-782, preparedby BWXT Yc12, represents the response to these RAIs.

Key issues raised in the RAIs concerned the technical basis for the proposed re-definition ofpyrophoric uranium. This response includes additional technical data to further back-up thedefinition. The final proposed definition is given on the attached ES-3 100 Safety AnalysisReport page 1- 12.

In addition, a mark-up of the CoC for this revision is included for your consideration. This mark-up represents CoC Revision 5, and should have used Revision 4 as the. base, but since thatrevision was not available at the time of this submittal, a mark-up of Revision 3 was used.

This submittal includes three attachments. The first is the RAI response document (Y/LF-782).The second is page changes to the ES-3100 SAR (Document No. Y/LF-717, RI). These affectedSAR pages are identified with a "page change 3" in the footer. A guide to insertion of these pagechanges in the SAR is included. Third is a mark-up of the CoC with changes directly applicableto the revised definition of pyrophonc uranium.

The applicant respectfully requests that these responses be reviewed on an expedited schedule. Arevised CoC is needed by mid-August.to support a Naval Reactors shipment of HEU brokenmetal in September. The original of this letter with an attachment is being sent to the DocumentControl Desk. Ten copies of this letter with attachments are being delivered to Kimberly J.Ha-din, Project Manager, Licensing Branch, Division of Spent Fuel Storage and Transportation,Office of Nuclear Material Safety and Safeguards.

0 h"'I'd ýCn ýoy 'A - -ýdýu p't"-z 0 frýS5O0

If you have any questions, please conLact me at 301-903-55 t3

Sincerely,

iT'

James ShulerManager, Packaging Certification ProgramSafety Management and OperationsOffice of Environmental Management

E•n cl os ureS.

cc:Kimberly J. Hardin, NRCJoe Bozik, NNSA NA-261Dana Willaford, DOE OROJeff Arbital, BWXT Y-12Steve Sanders, BWXT Y- 12

Docket No. 71-9315 RESPONSES TO RAI's on Revision 5 TAC No. L24063

ATTACHMENT 1

Y/LF-782

RAI RESPONSE DOCUMENT

ES-3100 Shipping ContainerY-12 National Security Complex

June 30, 2007Oak Ridge, TN

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TLBWXTV412A BWXT/Bechtel Enterprise

Y-12NATIONALSECURITYCOMPLEX

Y/LF-782

REQUEST FOR ADDITIONALINFORMATION

MODEL ES-3100 PACKAGE

Amendment 5Docket No. 71-9315

TAC No. L24063

Reference: Letter Kimberly J. Hardin,U.S. NRC to James M. Shuler, U.S. DOE

(dated May 9, 2007)

BWXT Y-12, L.L.C.

June 30, 2007

MANAGED BYBWXT Y-12, LLC.FOR THE UNITED STATESDEPARTMENT OF ENERGY

UCN-1 3672 (1 000)

Y/LF-782

REQUEST FOR ADDITIONAL INFORMATIONMODEL ES-3100 PACKAGE

Amendment 5Docket No. 71-9315

TAC No. L24063

Reference: Letter Kimberly J. Hardin, U.S. NRC to James M. Shuler, U.S. DOE(dated May 9, 2007)

June 30, 2007

Prepared by theY-12 National Security ComplexOak Ridge, Tennessee 37831

Managed byBWXT Y-12, L.L.C.

for theU.S. DEPARTMENT OF ENERGY

under Contract No. DE-AC05-84OR21400

RAI Response Document - Revision 5

1.0 General Information

Materials Review

1-1 Demonstrate that when considerations of the uncertainties in surface roughness,potential for hydride inclusions, packing densities, and measured ignitiontemperatures are evaluated, the maximum temperature reached by the canister staysbelow the lowest ignition temperature when uncertainties are considered, and thatthere is no pyrophoric potential for material that does not pass through the sieve.

Two curves in the Totemeier Studyl were used to propose ignition temperatures(340°C and the 5500C) for metal pieces with a surface area of I cm2/g. The plots didnot consider the purity of the gas or condition of the sample given. Ignitiontemperature is uncertain and is not an intrinsic value of the metal. The previoushistory and conditions of the samples used in these studies need to be described andcompared to the history of the samples intended to be shipped so that a safetydetermination can be made regarding these ignition temperatures.

This information is needed to determine compliance with 10 CFR 71.43 (d), 10 CFR71.55(d)(2), and 10 CFR 71.89.

Applicant Response:

Uncertainties in surface roughness, potential for hydride inclusions, and packing densitiesare addressed in response to other questions below. In addressing the uncertainty inignition temperature, the source reports for the Totemeier review article were consulted(Baker, Schnizlein, and Bingle,a and Tetenbaum, Mishler, and Schnizleinb), as well as anadditional report by Baker et al in the same journal.' These three, particularly the two byBaker et al, examine variations in the gas composition, in the uranium composition, and theuranium metallurgy.

A key aspect of the burning curves for various uranium compositions and metallurgies,burned in a range of oxygen concentrations, is the tendency for the uranium to"thermocycle." Once the uranium temperature reaches a critical value around 4000C thetemperature increases rapidly, then falls off, and then increases again. This is demonstratedin a number of graphs in the two Baker reports (see for. example figure 5 in footnote a).Baker, Schnizlein, and Bingle conclude (in both papers) that this is due to the transition from

a Journal of Nuclear Materials, Volume 20, 1966, L. Baker, L. G. Schnizlein, and J. D. Bingle, The Ignition ofUranium, pp 22-38.

b Nuclear Science and Engineering, Volume 14, 1962, M. Tetenbaum ,L. Mishler, and G. Schnizlein, UraniumPowder Ignition Studies, pp 230-238.

c Journal of Nuclear Materials, Volume 20, 1966, L. Baker, L. G. Schnizlein, and J. D. Bingle, The Ignition of BinaryAlloys of Uranium, pp 39-40.

1

Y/LF-782, Rev. 0, 06-30-07 Docket No. 71-9315; TAC No. L24063


an autocatalytic oxide layer to a protective layer as the uranium heats up through the 4000 to5000C range. In some cases, the transition is enough to dampen the first temperatureexcursion, while in other cases it is not. This difference - whether ignition occurs at the firsttemperature excursion or the changing nature of the oxide layer dampens the first rise andignition occurs later- is the source of the discontinuity in Baker (footnote a, figure 10), whichis reproduced in Totemeier figure 4. The lower curve in this figure, nearly straight across ata value of approximately 3900C, represents ignition occurring at this first temperatureexcursion in the burning curve. As specific surface areas drop, the change in the nature ofthe oxide layer is enough to dampen the first temperature excursion and delay ignition. AsBaker et al report, variations in metallurgy, metal composition, and gas composition couldvary the point at which ignition is delayed beyond the first temperature excursion, particularlyfor specific surface areas in the range of 2 to 6 cm 2/g.

Because variations in metallurgy, metal composition, and gas composition can causeignition to occur at the first temperature excursion (represented by the lower line in Totmeier,figure 4), the ES-3100 analysis used the value of 3900C from the lower line rather than5500C from the upper curve. This value corresponds well with the 3400C value reported inTetenbaum. A degree of conservatism was added by rounding these values down to 600 K(327 0C).

A footnote was added to the SAR text (page 1-204) with a brief explanation of this.

1-2 Revise the definition to state that shipment is limited to uranium metal that has notbeen stored in water. The operating instructions should include steps to ensure thiscondition is fulfilled.

According to the Totemeier review article, the ignition temperature is very dependenton the previous storage history, especially if the contents were in contact with water.As Totemeier indicates, if the samples had been exposed to a moist environment, UH3has been consistently found to form as a secondary phase with significantly differentignition characteristics. Depending on the conditions used to determine the ignitiontemperature and the past history of the uranium metal, restrictions on past historymight need be necessary.


Applicant Response:

The proposed definition on SAR pages 1-12 and 1-201 have been revised to state that ifuranium metal or alloy has ever been stored in water, it must be converted to an oxide orshipped in an inert gas atmosphere. The new section on operating procedures (SAR pages7-3 and 7-4) contains similar language.

2Y/LF-782, Rev. 0, 06-30-07 Docket No. 71-9315; TAC No. L24063


1-3 Demonstrate how Rho used in Equation 4 is a property of the can contents not theuranium metal.

Rho was used to convert surface reaction rates of the uranium to volumetric rates ofthe can contents in Equation 2. The reaction rates of the uranium metal should beused; otherwise Rho would have to be the density of the uranium fragment. Rho'sdefinition should not change from the density of the uranium to the smear density ofthe container.


Applicant Response:

To clarify the equations, the bulk density has been replaced with the metal density times apacking density, The requisite changes have been made in the SAR, specifically inequations 2 and 4, and in Table 1 (in SAR Appendix 1.4.10), as well as other placesthroughout SAR Appendix 1.4.10.

1-4 Revise, to be technically correct, the last line on p. 1-205 to read "maximum allowableuranium heat generation rate" as indicated in Equation 6.

Complete and accurate information must be provided in the application in accordancewith the 71.7(a) requirements.

Applicant Response:

The paragraph in question (directly preceding equation 6) has been reworded to say thatequation 6 calculates the "maximum possible uranium heat generation rate" from the

configuration being evaluated. This paragraph is now the first paragraph on SAR page1-206.

1-5 Show how the packing density affects the conclusion that the system is stableagainst ignition. Describe how a change to a density of 10 g/cm3 would affect theresults.

A bulk density of 5 g/cm3 is used on p. 1-207. This is only a 27% packing density.One fourth inch diameter beads should be able to be packed to a much higherdensity, probably 60-70%. The Epstein paper2 indicates that the ability to ignite, attemperatures even as low as ambient, depend on the.density and size of the bed ofmaterials.


3Y/LF-782, Rev: 0, 06-30-07 Docket No. 71-9315; TAC No. L24063


Applicant Response:

To show this affect, the bulk density has been replaced with the metal density times apacking density. A new section "Variation in Input Packing Density" has been added on SARpage 1-209 to examine a range of packing densities.

1-6 Interchange the phrases in the sentence in the definition starting with "Power..." onp. 1-201 to remove any ambiguity. The sentence should start with the phrase"Incidental small particles.., size restrictions," to clarify that wire, foils, etc., are neverpermitted to be transported unless they are pre-oxidized or contained in an inertatmosphere.


Applicant Response:

The sentence in question, in the proposed definition on the bottom of SAR page 1-201 hasbeen reworded for clarity, as requested by the reviewer.

1-7 Establish definitions to distinguish between a wire and a rod.

Definitions are needed to distinguish a rod from a wire. The SA/M criterion is only metif any long wire or rod has a diameter greater than 0.4 cm. A ball of tangled shavings,rods, and wires, all with a diameter < 0.4 cm, can be formed into a ball that will notpass through the proposed one fourth inch mesh, yet the SA/M will be larger than theallowable limit of I cm2/g. A proper definition of both a "wire" and a "rod" will assurethis ball is transported in an inert atmosphere.


Applicant Response:

Attachment 2 of SAR Appendix 1.4.10 (page 1-215) shows that rods of infinite length with adiameter of at least 1/4 inch (0.63 cm) would meet the <1 cm 2/g criteria. This item wouldpass through the 3/8-inch (0.95 cm) mesh and yet, if smooth, is clearly not pyrophoric. So,operators will be instructed to first evaluate the SA/M ratio, and if that can't be accomplished,their only recourse will be to use the 3/8-inch mesh screening process. This process hasbeen addressed in the new section regarding operating procedures on SAR pages 7-3 and7-4 (see also response to RAI 1-11). In this process, if the SAIM ratio can't be evaluated fora rod-shaped (or any other) item and it passes through the 3/8-inch mesh, it fails the test andmust be considered pyrophoric.

4



Wire (and by inference balls of wire) is specifically called out in the pyrophoric definition inthe CoC as requiring an inert atmosphere or be oxided prior to shipment. The operatingprocedures on SAR pages'7-3 and 7-4 expressly require wire and balls of wire to be treatedas pyrophoric. The operating procedures on SAR pages 7-3 and 7-4 allow rod-shaped orany other broken metal shape (defined on SAR page 1-12) to be evaluated to determine ifthey are pyrophoric.

A ball, or tangled mass, of small items (such as wires, rods, and turnings) will not be treatedas a solid shape. So even though the SA/M ratio can not be evaluated and the mass will notpass through the sieve, it will be treated as pyrophoric if any individual part of the mass ispyrophoric. The specialized operating procedures on SAR pages 7-3 and 7-4 address thetreatment of a tangled mass of small items in this manner.

1-8 Justify the roughness factor and be consistent in its use. Justification of the grid sizefor sieving should include a reasonable roughness factor for the metal.

Both examples on p. 1-202 should use the same roughness factor. Example #1currently has a roughness factor of I and example #2 uses 2. A factor of two in thefirst example would drive the SA/M over the allowable limit of I cm2/g. Additionally,provide a justification for using the factor of 2. Most surface roughness factors rangefrom 3 to 6.


Applicant Response:

The roughness factor of two was based on a recommendation from Y-12 plant personnel.in response to this question, additional evaluations were made of fracture surfaces of casturanium metal. Observed roughness factors ranged from 1.1 to 2.7, with a mean value of2.0. This value is consistent with the recommendation from Y-1 2 experience, but since themaximum observed value was larger than 2, a roughness factor of three was selected tobound the highest value observed.. The previous % inch mesh definition was changed to 3/8inch mesh throughout the text. A discussion of this was added to SAR pages 1-210 and1-211. Example 1 (SAR page 1-202) was modified to compare the results for a smoothsurface and a rough surface, and example 2 was deleted.

1-9 Revise the proposed definition of pyrophoricity on p. 1-201 to state: "if an inertatmosphere is used, the cover gas will have a minimum of 99.999% purity," or includea section that warns the consignee that the convenience cans must be opened in aninert cell to avoid possible pyrophoric events. Include these in the operatingprocedures.

Totemeier indicated that for uranium metal sealed in an inert atmosphere, there was

5



sufficient moisture in the cover gas to produce UH3 and a U02 surface layer. Inaddition, studies by Solbrig3 indicate that not all the hydrogen produced in theuranium/moisture reaction forms UH3 so there is free hydrogen in the canister; asmuch as 13%. When air is admitted to the canister, the 2UH3 + 202 2UO2 + 3H2reaction can occur. This is a rapid and exothermic reaction. Totemeier states that;"The heat generated could either ignite any hydrogen gas present in the container orthe uranium itself." Totemeier also discusses potential ways to avert this situationand the pros and cons of each.


Applicant Response:

The revised definition of pyrophoricity on SAR page 1-201 states "under an inert cover gas."In the new section on operating procedures (SAR pages 7-3 and 7-4), the cover gasspecifications are given as argon with high purity (> 99.997%) and dry (< 5 ppm moisture).These values are Y-12's standard purchase specifications for argon, and sufficient to

eliminate hydride formation.

1-10 Clarify in Chapter 7 what steps are being taken to assure that the fines have beensufficiently oxidized so water vapor does not diffuse through the oxide layer to themetal underneath and establish the potential for hydride formation.

The operating instructions for pre-transport oxidation of fines (wires, etc.) should berevised to state the conversion of the metal to an oxide prior to transport must becomplete enough to prevent diffusion of water vapor through the oxide to the metalsurface over the maximum period of transport, i.e., one year.


Applicant Response:

Any small particles which are oxidized (rather than packed in an inert atmosphere) will be"completely oxidized." This is noted in the revised definition of pyrophoricity, the new sectionregarding operating procedures on SAR pag'es 7-3 and 7-4, and in SAR Appendix 1.4.10 onpages 1-211 and 1-212.

1-11 Include a section describing operator training and guidance for doing the sieving thatis important to separating the size factions of the contents.


6Y/LF-782, Rev. 0, 06-30-07 Docket No. 71-9315-, TAC No. L24063


Applicant Response:

A new section has been included in Section 7, Package Operations, which include operatingprocedures for overall determination of pyrophority before loading the ES-3100 (SAR pages7-3 and 7-4). This information is also in SAR Appendix 1.4.10 on pages 1-211 and 1-212.

References

1. Totemeier, Terry C., ANL/ED/95-2, "A Review of the Corrosion and PyrophoricityBehavior of Uranium and Plutonium," June 1995, Argonne National Lab-West.

2. M. Epstein, B. Malinovic, M.G. Plys, SNF-6192-FP, "Uranium Pyrophoricity Phenomenaand Prediction," April 2000, Flour Hanford, Richland, Washington.

3. C.W. Solbrig, J.R. Krsul, and D.N. Olsen, "Pyrophoricity of Uranium in Long-term StorageEnvironments," 1994, DOE Spent Fuel Challenges and Initiatives, American NuclearSociety.

Y/LFJ82, Rev. 0,, 06-30-07 Docket No. 71-9315; TAC No. L24063


ATTACHMENT 2

SAR PAGE CHANGES




GUIDE TO PAGE CHANGES

Y/LF-717, Rev. 1, page change #3

SAR SECTION PAGE CHANGES,

Volume 1, Front section Replace pages i to xx

Volume 1, Section 1 Replace pages 1-11 to 1-12

Replace pages 1-199 to 1-218

Volume 2, Front section Replace pages i to xx

Volume 2, Section 6 Replace page 6-239

(page numbering fix from page change #2)

Volume 2, Section 7 Replace pages 7-1 to 7-14



Y/LF-717, Rev. .1, Page g•je3Volume I

SAFETY ANALYSIS REPORT,Y-12 NATIONAL SECURITY COMPLEX,

MODEL ES-3100 PACKAGE WITH BULK HEU CONTENTS

Prepared by theOak Ridge Y- 12 National Security Complex

Oak Ridge, Tennessee 37831Managed by

BWXT Y-12, L.LC.for the

LI. S. Department of Energyunder contract DE-AC05-840R21400

Junie 3'O,-2'007

Y/LF-717/Rev I/page change 3nlro/ES-3100 HEU SAR/sc/6-30-07

0

ii

Y/LF-717/Rev I/page change 3/lntro/ES-3 100 HEU SAR/sc/6-30-07

CONTENTS

L IST O F FIG U R ES .................................................................. xi

LIST O F T A B LES .................................................................. xiii

ACRONYMS AND ABBREVIATIONS ............................................... xvii

R EV ISIO N LO G ................................................................... xix

1. G EN ERAL IN FORM ATION ...................................................... 1-1

1.1 IN TRO D U C TIO N .......................................................... 1-1

1.2 PACKAGE DESCRIPTION .................................................. 1-11.2.1 Packaging .................................................... 1-41.2.2 Containm ent System ................................................... 1-71.2.3 C ontents ........................................................... 1-91.2.4 O perational Features ................................................. 1-.19

1.3 GENERAL REQUIREMENTS FOR ALL PACKAGES ...... 1-191.3.1 M inim um package size ............................................... 1-191.3.2 Tamper-indicating feature ........................................ 1-20

1.4 APPENDICES ............................ ..... ..................... 1-211.4.1. CERTIFICATION DRAW ING ........................................ 1-231.4.2 EQUIPMENT SPECIFICATION JS-YMN3-801580-A002,

ES-3100 DRUM ASSEMBLY .......................................... 1-371.4.3 EQUIPMENT SPECIFICATION JS-YMN3-801580-AOOI,

ES-3100 CONTAINMENT VESSEL ..................................... 1-471.4.4 EQUIPMENT SPECIFICATION, JS-YMN3-801580-A003,

MANUFACTURING PROCESS SPECIFICATION FOR CASTINGKAOLITE 1600TM INTO THE ES-3100 SHIPPING PACKAGE ............... 1-61

1.4.5 EQUIPMENT SPECIFICATION, JS-YMN3-801580-A005, CASTINGCATALOG NO. 2 77-4 NEUTRON ABSORBER FOR THE ES-3100SHIPPING PA CKA GE ............................................... 1-81

1.4.6 PACKAGE CATEGORY DETERMINATION ........................... 1-1211.4.7 HEU OXIDE MATERIAL SPECIFICATION AS PROVIDED BY Y-12

HIGHLY ENRICHED URANIUM DISPOSITION PROGRAM OFFICE ...... 1-1251.4.8 DETAILED ENGINEERING DRAWINGS ............................. 1-1291.4.9 DESIGN ANALYSES AND CALCULATIONS, DAC-PKG-801624-A001,

MIXING WEIGHTS AND ELEMENTAL COMPOSITION OF 277-4NEUTRON POISON USED IN THE ES-3100 .......................... 1-171

Vf-'4.1,O PYRopHORICIt, OF)'ANIUMM"ETAL: .......................... 1-199

SECTION 1 REFERENCES ......................................................... l-21

iii

Y/LF-717/Rev I/page change 3/lntro/ES-3100 HEU SAR/sc/06-30-07

CONTENTS continued

2. STRUCTURAL EVALUATION................................................. 2-1

2.1 DESCRIPTION OF STRUCTURAL DESIGN ................................... 2-22.1.1 Discussion ............................................. I........... 2-22.1.2 Design Criteria .................................................... 2-42.1.3) Weights and Centers of Gravity ...................................... 2-142.1.4 Identification of Codes and Standards for Package Design .................. 2-14

2.2 MATERIALS.......................................................... 2-182.2.1 Material Properties and Specifications ................................. 2-182.2.2, Chemical, Galvanic, or Other Reactions ................................ 2-182.2.3- Effects of Radiation on Materials ..................................... 2-26

2.3 FABRICATION AND EXAMINATION ....................... .............. 2-262.3.1 Fabrication ..................................................... 2-262.3.2 Examination ..................................................... 2-28

2.4 LIFTING AND TIE-DOWN STANDARDS FOR ALL PACKAGES ................. 2-312.4.1 Lifting Devices ................................................... 2-3 12.4.2 Tie-Down Devices .. ............................................. 2-31

2.5 GENERAL CONSIDERATIONS ........................................... 2-322.5.1 Evaluation by Test.......................................... ...... 2-322.5.2 Evaluation by Analysis............................................. 2-35

2.6 NORMAL CONDITIONS OF TRANSPORT .................................. 2-352.6.1 Heat........................................................... 2-372.6.2 Cold............................................. ..... 2-392.6.3 Reduced External Pressure.......................................... 2-402.6.4 Increased External Pressure ......................................... 2-402.6.5 Vibration........................................................ 2-442.6.6 Water Spray..................................................... 2-462.6.7 Free Drop ...................................................... 2-472.6.8 Corner Drop........................................... **'*'*****..... 2-472.6.9 Compression,.................................................... 2-482.6.10 Penetration ..................................................... 2-50

2.7 HYPOTHETICAL ACCIDENT CONDITIONS................................. 2-512.7.1 Free Drop....................................................... 2-532.7.2 Crush.......................................................... 2-642.7.3 Puncture........................................................ 2-702.7.4 Thermal........................................... 2-712.7.5 Immersion-Fissile Material......................................... 2-792.7.6 Immersion-All Packages ......... I................................. 2-812.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2) 2-812.7.8 Summary of Damage.............................................. 2-81

2.8 ACCIDENT CONDITIONS FOR AIR TRANSPORT OF PLUTONIUM .............. 2-95

iv

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CONTENTS continued

2.9 ACCIDENT CONDITIONS FOR FISSILE MATERIAL PACKAGES FOR AIRTRANSPORT ..................... ................................. 2-95

2.9a SPECIA L FO RM ........................................................... 2-95

2.9b FU EL RO D S ............................................................. 2-95

2.10 APPENDICES ...................................................... 2-972.10.1 ES-3 100 CONTAINMENT VESSEL ASME CODE EVALUATION

(DAC-EA-900000-A006 and DAC-EA-900000-A007) ...................... 2-992.10.2 IMPACT ANALYSES OF ES-3 100 DESIGN CONCEPTS USING

BOROBOND AND CAT 277-4 NEUTRON ABSORBERS ............... 2-1572.10.3 KAOLITE PROPERTIES ............................................ 2-4632.10.4 CATALOG 277-4 PROPERTIES ..................................... 2-5352.10.5 BOROBOND4 PROPERTIES ......................................... 2-6432.10.6 RECOMMENDED RANDOM VIBRATION AND SHOCK TEST

SPECIFICATIONS FOR CARGO TRANSPORTED ON SST AND SGTT RA ILER S ....................................................... 2-657

2.10.7 TEST REPORT OF THE ES-3 100 PACKAGE, VOLUME 1 - MAIN REPORT,ORNL/NTRC-013/VI, REV. 0, SEPTEMBER 10, 2004 ................... 2-675

2.10.8 TEST REPORT OF THE ES-3 100, VOLUME 3 - APPENDIX K, TU - DATASH E ET S ......................................................... 2-807

210•.)9 PACKAGING MATERIALS OUTGASSI S DYFNA REPORT ..... 2-851

SECTION 2 REFERENCES ............................. ...................... 2-891

3. THERMAL EVALUATION. ............................. 3-1

3.1 D ISC U SSIO N .............................................................. 3-13.1.1 D esign Features ................................ ..................... 3-33.1.2 Content's Decay Heat ................................................. 3-43.1.3 Summary Tables of Temperatures ................................... 3-43.1.4 Summary Tables of Maximum Pressures ................................ 3-11

3.2 SUMMARY OF THERMAL PROPERTIES OF MATERIALS .................... 3-163.2.1 Material properties ............................................. 3-163.2.2 Component Specifications ........................................... 3-16

3.3 GENERAL CONSIDERATIONS ............................................ 3-163.3.1 Evaluation by A nalysis .................................. ............ 3-163.3.2 Evaluation by Test ............. .... ....................... ......... 3-183.3.3 M argins of Safety .................................................. 3-23

3.4 THERMAL EVALUATION UNDER NORMAL CONDITIONS OF TRANSPORT .... 3-273.4.1 Heat and Cold ............................................. .... 3-273.4.2 Maximum Normal Operating Pressure .......................... 3-293.4.3 M aximum Thermal Stresses .......................................... 3-29

Y/LF-717/Rev 1/page change 3/intro/ES-3100 HEU SAR/sc/06-30-07

CONTENTS continued

3.5 HYPOTHETICAL ACCIDENT THERMAL EVALUATION ...................... 3-313.5.1 Initial Conditions .............................................. 3-313.5.2 Fire Test conditions ............................................ 3-313.5.3 Maximum Temperatures and Pressure ................................... 3-333.5.4 Accident Conditions for Fissile Material Packages for Air Transport .......... 3-38

3.6 A PPEN D IC ES ........................................................... 3-393.6.1 THERMAL EVALUATION OF THE ES-3 100 SHIPPING CONTAINER

FOR NCT AND HAC (CONCEPTUAL DESIGN WITH BOROBOND4NEUTRON ABSORBER) ......................................... 3-41

3.6.2 THERMAL EVALUATION OF THE ES-3 100 SHIPPING CONTAINERFOR NCT AND HAC (FINAL DESIGN WITH CATALOG 277-4NEUTRON ABSORBER)........................................ 3-81

3.6.3 THERMAL STRESS EVALUATION OF THE ES-3 100 SHIPPINGCONTAINER DRUM BODY ASSEMBLY FOR NCT (FINAL DESIGNWITH CATALOG 277-4 NEUTRON ABSORBER) .................... 3-121

3.6.4 CONTAINMENT VESSEL PRESSURE DUE TO NORMAL CONDITIONSOF TRANSPORT FOR THE PROPOSED CONTENTS .................. 3-145

3.6.5 CONTAINMENT VESSEL PRESSURE DUE TO HYPOTHETICALACCIDENT CONDITIONS FOR THE PROPOSED CONTENTS .......... 3-153

3.6.6 SILICONE RUBBER THERMAL PROPERTIES FROMTHERM 1.2 DATABASE .......................................... 3-159

SECTION 3 REFEREN CES ......... ........................... .................... 3-161

4. CON TA IN M EN T .............................................................. 4-1

4.1 DESCRIPTION OF THE CONTAINMENT BOUNDARY ......................... 4-24.1.1 Containment Boundary ........................................... 4-24.1.2 Special Requirements for Plutonium .................................... 4-4

4.2 GENERAL CONSIDERATIONS .............................. ...... 4-44.2.1 Type A Fissile Packages ................................................ 4-44.2.2 Type B Packages ............................................... 4-4

.4.3 CONTAINMENT UN DER NORMAL CONDITIONS OF TRANSPORT(TY PE B PACKA G ES) ...................................................... 4-6

4.4 CONTAINMENT UNDER HYPOTHETICAL ACCIDENT CONDITIONS

(TY PE B PACKAG ES) ..................................... ................ 4-8

4.5 LEAKAGE RATE TESTS FOR TYPE B PACKAGES ......................... 4-10

4.6 A PPEN D IC ES .............. ............................................ 4-114.6.1 DETERMINATION OF A, FOR THE ES-3 100 PACKAGE WITH HEU

C O N T EN T S ...................................................... 4-134.6.2 CALCULATION OF THE ES-3 100 CONTAINMENT VESSEL'S

REGULATORY REFERENCE AIR LEAKAGE RATES .................. 4-23

SECTION 4 REFERENCES ......................................................... 4-35

vi

Y/LF-717/Rev 1 /page change 3/Intro/ES-3 100 HEU SAR/sc/06-30-07

CONTENTS continued

5. SHIELDING EVALUATION ................................................ 51

5.1 DESCRIPTION OF SHIELDING DESIGN ..................................... 5-15.1.1 Design Features ................................................ 5-15.1.2 Summary Table of Maximum Radiation Levels ............................ 5-1

5.2 SOURCE SPECIFICATION .................. .................... ....... 5-I

5.3 DOSE RATE ANALYSIS MODELS .......................................... 5-35.3.1 Packaging Model Conservative Features ................................. 5-75.3.2 Photon model for 36-kg HEU metal content .............................. 5-95.3.3 Neutron model for 36-kg HEU metal content ........................... 5-115.3.4 Photon model for 24-kg HEU oxide content ............................. 5-115.3.5 Neutron model for 24-kg HEU oxide content ............................. 5-11

5.4 SHIELDING EVALUATION ................................................ 5-11

5.5 APPENDICES ...................................................... 5-155.5.1 ORIGEN INPUT DATA FROM TABLE 5.3 ............................ 5-175.5.2 CSASN AND ICE INPUT FROM TABLE 5.8 ........................... 5-195.5.3 MORSE ROUTINES AND INPUT DATA ............................... 5-21


6. CRITICALITY EVALUATION ................................................... 6-1

6.1 DESCRIPTION OF THE CRITICALITY DESIGN ............................... 6-16.1.1 Design Features ............................................ ........ 6-16.1.2 Summary of the Criticality Evaluation ................................... 6-26.1.3 Criticality Safety Index ........................ ....................... 6-4

6.2 PACKAGE CONTENTS .................... .............................. 6-306.2.1 Fissile Material Contents .................................. ....... 6-306.2.2 Convenience Cans, Teflon and Polyethylene Bottles, and 277-4 Canned Spacers 6-316.2.3 Packing Materials .............................................. 6-316.2.4 Package Content Loading Restrictions .................................. 6-32

6.3 GENERAL CONSIDERATIONS ............................................ 6-336.3.1 M odel Configuration ................................................ 6-346.3.2 M aterial Properties .......... ...................................... 6-506.3.3 Computer Codes and Cross-Section Libraries ............................ 6-616.3.4 Demonstration of Maximum Reactivity .............................. 6-63

6.4 SINGLE PACKAGE EVALUATION ......................................... 6-646.4.1 Solid HEU Metal of Specified Geometric Shapes ......................... 6-656.4.2 HEU Solid Metal of Unspecified Geometric Shapes or HEU Broken Metal .... 6-686.4.3 HEU Oxide ....................................................... 6-686.4.4 U N H C rystals ...................................................... 6-70

vii


CONTENTS continued 06.5 EVALUATION OF PACKAGE ARRAYS UNDER NORMAL

CONDITIONS OF TRANSPORT ........................................ 6-716.5.1 Solid HEU Metal of Specified Geometric Shapes ......................... 6-7.16.5.2 HEU Solid Metal of Unspecified Geometric Shapes or HEU Broken Metal ..... 6-746.5.3 HEU Oxide .................................................. 6-756.5.4 UN H Crystals ..................................................... 6-75

6.6 EVALUATION OF PACKAGE ARRAYS UNDER HYPOTHETICALACCIDENT CONDITIONS ............................................ 6-766.6.1 Solid HEU Metal of Specified Geometric Shapes ......................... 6-766.6.2 HEU Solid Metal of Unspecified Geometric Shapes or HEU Broken Metal ..... 6-776.6.3 H EU O xide ....................................................... 6-786.6.4 UNH Crystals ................................................ 6-78

6.7 FISSILE MATERIAL PACKAGES FOR AIR TRANSPORT ...................... 6-796.7.1 Results for Solid HEU, One Piece per Convenience Can .................... 6-796.7.2 Results for TRIGA Fuel Elements, Three Pieces per Convenience Can ........ 6-836.7.3 Results for HEU Broken Metal, More Than One Piece per Convenience Can ... 6-856.7.4 Conclusions ................................................. 6-87

6.8 BENCHMARK EXPERIMENT'S ............................................ 6-876.8.1 Applicability of Benchmark Experiments ............................. 6-876.8.2 Details of Benchmark Calculations ...... ............................. 6-876.8.3 Bias Determination ............................................. 6-88 0

6.9 APPENDICES ....................................................... 6-896.9.1 FISSILE CONTENT MODELS .................................... 6-916.9.2 HAC PACKAGE MODEL ....................................... 6-976.9.3 PACKAGE MATERIAL COMPOSITIONS ............................ 6-1096.9.4 QUALIFICATION OF A NEUTRON ABSORBER MATERIAL

FOR THE ES-3 100 ............................. ........ 6-1396.9.5 MISCELLANEOUS INFORMATION AND DATA.................... 6-1516.9.6 ABRIDGED SUMMARY TABLES OF CRITICALITY

CALCULATION RESULTS ..................................... 6-1556.9.7 INPUT LISTINGS OF ES-3 100 CALCULATION MODELS FOR

SELECT CASES IDENTIFIED IN TABLES 6. la-e ...................... 6-269

SECTION 6 REFERENCES ........................................................ 6-445

7. PACKAGE OPERATIONS ....................................... ........... 7-1

7.1 PACKAGE LOADING ................................................ 7-17.1.1 Preparation for Loading .......................................... 7-17.1.2 Loading of Contents ..................................... i ........... 7-37.1.3 Preparation for Transport ......................................... 7-7

7.2 PACKAGE UNLOADING ............ .................. .................. 7-97.2.1 Receipt of Package from Carrier ........................................ 7-97.2.2 Removal of Contents ........................................... 7-11

viii

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CONTENTS continued

7.3 PREPARATION OF EMPTY PACKAGE FOR TRANSPORT ..................... . -li

7.4 OTHER OPERATIONS .................................................... .7-

SECTION 7 REFERENCES .....................................................- 3

8. ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ........................... 8-1

8.1 ACCEPTANCE TESTS ..................................................... 8-18.1,1 Visual Inspections and Measurements .............. ...................... 8-38.1,2 W eld Exam inations ................................................... 8-48.1.3 Structural and Pressure Tests ........................................... 8-58.1.4 Leakage Tests ....................................................... 8-58.1.5 Component and M aterial Tests .......................................... 8-68.1L 6 Shielding Tests ........ .............................................. 8-68.1,7 Therm al Tests ....................................................... 8-68.1.8 M iscellaneous Tests .................................................. 8-7

8.2 MAINTENANCE PROGRAM ................................................ 8-78.2.1 Structural and Pressure Tests .......................................... 8-78.2.2 Leakage Tests ....................................................... 8-78.2.3 Component and M aterial Tests .......................................... 8-88.2.4 Thermal Tests ............................................. 8-98.2,5 M iscellaneous Tests ......... ........................................ 8-9

SECTION 8 REFERENCES ..................................................... 8-11

D IST R IB U T IO N .................................................................... 8-13

ix

Y/LF-717/Rev 1/page change 3/Intro/ES-3100 HEU SAR/sc/06-30-07

LIST OF FIGURES

1.1 Schematic of the ES-3 100 shipping package ........................................ 1-21.2 Exploded view of the ES-3 100 package with bulk HEU contents ........................ 1-31.3 Containment boundary of the ES-3 100 shipping package .............................. 1-81.4 Typical shipping configurations inside the ES-3 100 containment vessel ................. 1-15

1.5 fuel lem ent . .. . .......... .... ....... . . .......2.1 Containment vessel calculated stress locations .................... ................ 2-112.2 ES-3 100 shipping package center of gravity locations ................................ 2-152.3 ES-3 100 vibration testing arrangement ........................................... 2-442.4 W ater spray test arrangement for Test Unit-4 ...................................... 2-452.5 NCT free drop test on Test Unit-4 ................................................ 2-472.6 Compression test on Test Unit-4 ................................................ 2-482.7 Penetration test damage on Test Unit-4 ............................................. 2-492.8 9-m drop test arrangement for all test units ........................................ 2-532.9 9-m drop test damage on Test Unit-4 ............................................. 2-552.10 Cumulative damage from 9-m drop and crush testing on Test Unit-2 ................... 2*572.11 Test Unit-3 damage from 1.2 and 9-m drop tests .................................... 2-582.12 1.2 and 9-m drop test damage on Test Unit-I ...................................... 2-602.13 1.2 and 9-m drop test damage to Test Unit-5 ....................................... 2-622.14 Cumulative damage following 9-m crush on Test Unit-I ............................. 2-642.15 Cumulative damage following 9-m crush test on Test Unit-3 ................ ....... 2-662.16 Cumulative damage from 9-m drop and crush testing on Test Unit-4 ................... 2-672.17 Cumulative damage from 9-m drop and crush testing on Test Unit-5 .................... 2-692.18 28' oblique and horizontalpuncture tests on Test Unit-I ............................. 2-712.19 40' oblique puncture test on Test Unit-I ......................... ................ 2-712.20 Horizontal puncture test over Test Unit-I's containment vessel flange ................... 2-722.21 Horizontal CG puncture test on Test Unit-2 .................................... 2-722.22 24.60 oblique puncture test on Test Unit-3 ........................................ 2-732.23 Vertical puncture test on Test Unit-4 ............................................. 2-732.24 Horizontal puncture test over Test Unit-5's containment vessel flange ................... 2-742.25 Visual comparison of the cumulative damage on the. crush side surface after the three drop

tests (from top to bottom: Test Unit-I, analytical results with BoroBond, analytical resultswith Cat 277-4) ........................................................ 2-83

2.26 Visual comparison of the cumulative damage on the rigid surface side after the four droptests (from left to right: Test Unit-2, analytical results with BoroBond, analytical resultswith Cat 277-4) ................ 2-86

2.27 Visual comparison of the cumulative damage on the crush plate side after the three droptests (from left to right: Test Unit-2, analytical results with BoroBond, analytical resultsw ith C at 277-4) ................... .......................................... 2-86

2.28 Visual comparison of the cumulative bottom damage after the three drop tests (from left toright: Test Unit-3. analytical results with BoroBond, analytical results with Cat 277-4) ..... 2-88

2.29 Visual comparison of the cumulative lid damage after the three drop tests (from top tobottom: Test Unit-3, analytical results with BoroBond, analytical results with Cat 277-4) ... 2-89

2.30 Visual comparison of the cumulative damage after the three drop tests (from left to right:Test Unit-4, analytical results with BoroBond, analytical results with Cat 277-4) .......... 2-91

2.31 Containment vessel markings at assembly (swivel hoist ring removed prior to testing) ...... 2-932.32 Containment vessel marking after compliance testing ................................ 2-943.1 MSC.Patran axisymmetric finite element model of the ES-3 100 shipping container

with BoroBond4-nodal locations of interest (elements representing air not shownfor clarity ) . ................................................ ................... 3-6

xi

Y/LF-717/Rev I/page change 3/tntro/ES-3 100 HEU SAR/sc/06-30-07

LIST OF FIGURES continued

3.2 Test unit preheat arrangement ................................................... .3-24-23.3 Test unit insertion into furnace . ................................................. 3-243.4 Test unit removal from furnace.............................................. 3-253.5 Test unit cool down and monitoring arrangement .................................... 3-255.1 Cylindrical calculational model of the ES-3 100 shipping package for NCT ................ 5-65.2 ES-3 100 HEU metal content radial (top view) geometric models ....................... 5-106.1. R/Z section view of ES-3 100 single-unit packaging model .......................... 6-356.2. R/Z section view at bottom of ES-3 100 single-unit packaging showing

KENO V.a geometry units 100 1-1003, and 1006 (partial) .......................... 6-366.3. R/Z section view at center of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1006 (partial) ......................................... 6-376.4. R/Z section view of near top of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1008 (partial) and 1010-1016 ............................. 6-386.5. R/Z section view at the top of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1016-1019 ........................................... 6-396.6. R/Z section view of ES-3 100 array packaging model ................................ 6-416.7. R/Z section view at the bottom of the ES-3 100 array packaging showing

KENO V.a geometry units 1001-1003, and 1006 (partial) .......................... 6-426.8. R/Z section view at the center of the ES-3 100 array packaging showing

KENO V.a geometry units 1006 (partial), 1007, and 1008 (partial) .................... 6-436.9. R/Z section view of near top of the ES-3 100 array packaging showing,

KENO V.a geometry units 1008 (partial) and 1010-1016 .............................. 6-446.10. R/Z section view at the top of the ES-3 100 array packaging showing

KENO V.a geometry units 1016-1019 ............................................ 6-456 .11. M odel I . ...... .. .. ..... .................................... ............. ... 6-4 96 .12 . M odel 2 . ...... ............................................................. 6-496.13. Model3 ............................................................. 6-496.14. Model 4. ............................................................. 6-506.15. M odel 5 . ............................................................ 6-506 .16 . M odel 6 . ................................................................... 6-506.16a. ýKe + 20 versus pitch for triangular arrangeient of content in CV ...... ...... 6-66a6.16b. Ke, + 2a versus triangular pitch over range of CV mderation.... 6-66a6:16c.K l e +!2c2versus as a function of uranium we.ghtff.actoniandmoderation 6•-•66.17. K,:# vs. enrichment (wt % 235U) for HEU ranging from 7 to 10 kg 6-816.18. K,fl, vs. stainless steel shell radius (cm) for 7 kg HEU with enrichments ranging from

19 tO 100 wt % 2 35 U ..................................................... 6816.19. K,,, vs. Kaolite shell radius (cm) for 7 kg HEU with enrichments ranging from

19 to 100 wt % 235 U ........................................... 6-826.20. K,,, vs. shell radius :(cm) for dry and water-saturated Kaolite for 7 kg HEU at

100% enrichm ent ............................................................. 6-826.21. K,,, vs shell radius (cm) for 10.4 kg core of UZrH\ with stainless steel or Kaolite shell ...... 6-836.22. K,,,. vs. core radius (cm)'for homogenized core of UZrH,, 500 g polyethylene, and Kaolite

where the Kaolite water content ranges from the dry to the water-saturated condition ...... 6-846.23. K,., vs. excess water from Kaolite for fissile core blanketed with a variable thickness

Kaolite shell .................................. 6-856.24. K,,y.vs. excess water from Kaolite for core of 1-7 kg enriched HEU broken metal at

100% enrichment, core blanketed with a variable thickness Kaolite shell ................ 6-866.25. K,,,vs. excess water from Kaolite for core 20% enriched HEU broken metal

(0.6-7 kg 231U), core blanketed with a variable thickness Kaolite shell .................. 6-86

xii


LIST OF TABLES

1.1 Uranium concentration lim its ................ .. ................................ 1-101.2 Bounding uranium isotopic concentrations in oxide ............................... 1-101.3 Authorized content and fissile mass loading limits for ground transport ................. 1-131.4 TRG 'Fuel Spcifications .... 1-20a

2.1 Proposed HEU contents for shipment in the ES-3 100 * ................................. 2-12.2 Category designations for Type B packages ..................................... 2-72.3 Summary of load combinations for normal and hypothetical accident conditions of transport 2-92.4 Containment vessel allowable stress .............................................. 2-102.5 Allowable stress intensity (S m) for the containment boundary construction materials of

construction ................................................................ 2-102.6 ES-3 100 containment boundary design evaluation allowable stress comparisons .......... 2-122.7 ES-3 100 packaging material specifications ..................................... 2-132.8 Packaging weights for various ES-3 100 shipping package arrangements ................ 2-162.9 Compliance test unit weights .............................................. 2-172.10 Calculated center of gravity for the various ES-3 100 shipping arrangements .............. 2-182.11 Applicable codes and standards for Category I packaging ............................ 2-202.12 Mechanical properties of the metallic components of the drum assembly ......... 2-212.13 Mechanical properties of the lid fastening components for the drum assembly ............ 2-222.14 Mechanical properties of the cast refractory insulation ............................... 2-232.15 Mechanical properties of containment vessel 0-rings .............................. 2-232.16 Mechanical properties of the metallic components of the containment boundary ........... 2-242.17 Mechanical properties of the cast neutron absorber .................................... 2-252.18 Summary of NCT - I OCFR71.71 tests for ES-3 100 package ... ..................... 2-332.19 Summary of HAC - I 0CFR71.73 tests for ES-3 100 package ........................ 2-332.20 Summary of temperatures and pressures for NCT ................................... 2-362.21 ES-3 100 containment boundary evaluation for both hot and cold conditions .............. 2-412.22 NCT ES-3 100 containment boundary stress compared to the allowable stress

at reduced and increased external pressures ........................................ 2-422.23 Test and analysis summary for the ES-3 100 package .............................. 2-512.24 Recorded height damage to Test Unit-4 from 1.2-m and 9-m drop testing ................ 2-542.25 Recorded diametrical damage to Test Unit-4 from 1.2-m and 9-m drop tests .............. 2-542.26 Recorded diametrical damage to Test Unit-2 from NCT and HAC drop testing............ 2-562.27 Recorded flat contour damage to Test Unit-2 from NCT and HAC drop testing ........... 2-562.28 Recorded height damage to Test Unit-2 from NCT and HAC drop testing ................ 2-562.29 Recorded height damage to Test Unit-3 from 1.2-m and 9-m drop testing ................ 2-582.30 Recorded diametrical damage to Test Unit-3 from 1.2-m and 9-m drop testing ............ 2-582.31 Recorded height damage to Test Unit-I from 1.2-m and 9-m drop testing ................ 2-592.32 Recorded diametrical damage to Test Unit-I from 1.2-m and 9-m HAC drop testing ........ 2-592.33 Recorded flat contour damage to Test Unit-I from 1.2-m and 9-m drop testing ............ 2-602.34 Recorded height damage to Test Unit-5 from 1.2-m and 9-m drop testing ......... ...... 2-612.35 Recorded diametrical damage to Test Unit-5 from 1.2-m and 9-m HAC drop testing ....... 2-612.36 Recorded flat contour damage to Test Unit-5 from 1.2-m and 9-m drop testing ............ 2-622.37 Recorded height damage to Test Unit-I from the 9-m crush test ........................ 2-632.38 Recorded diametrical damage to Test Unit-I from the 9-m crush test ............ ........ 2-632.39 Recorded flat contour damage to Test Unit-I from the 9-m crush test .................. 2-642.40 Recorded height damage to Test Unit-3 from the 9-m crush test ...................... 2-652.41 Recorded flat contour damage to Test Unit-3 from the 9-m crush test ................... 2-652.42 Recorded diametrical damage to Test Unit-3 from the 9-m crush test ................... 2-65

xiii


LIST OF TABLES continued

2.43 Recorded height damage to Test Unit-4 from the 9-m crush test ...................... 2-662.44 Recorded diametrical damage to Test Unit-4 from the 9-m crush test .................... 2-672.45 Recorded height damage to Test Unit-5 from the 9-m crush test ........................ 2-682.46 Recorded diametrical damage to Test Unit-5 from the 9-m crush tests ................... 2-682.47 Recorded flat contour damage to Test Unit-5 from the 9-m crush test ................... 2-682.48 1-rn (40-in.) puncture drop test description and results .... ......................... 2-702.49 Thermax temperature indicating patches for test units ................................ 2-742.50 Maximum HAC temperatures recorded on the test packages' interior surfaces ............ 2-772.51 HAC ES-3 100 containment boundary stress compared to the allowable stress ............ 2-792.52 Diametrical damage comparison of Test Unit-I with analytical predictions ............... 2-822.53 Flat contour damage comparison of Test Unit-I with analytical results .................. 2-822.54 Cumulative analytical 120 slapdown drop tests maximum effective plastic strain results .... 2-842.55 Diametrical damage comparison of Test Unit-2 with analytical predictions .............. 2-852.56 Flat contour damage comparison of Test Unit-2 with analytical predictions .............. 2-852.57 Cumulative analytical side drop test maximum effective plastic strain results ............. 2-872.58 Diametrical damage comparison of Test Unit-3 with analytical predictions ............. 2-882.59 Cumulative analytical corner drop test maximum effective plastic strain results ........... 2-902.60 Diametrical damage comparison of Test Unit-4 with analytical predictions .............. 2-912.61 Cumulative analytical top drop test maximum effective plastic strain results ............. 2-922.62 ES-3 100 test package weights before and after 10 CFR 71.73(c)(4) HAC thermal testing .... 2-923.1 Isotopic mass and weight percent for the HEU contents ............................... 3-43.2 Decay heat for 35.2 kg of HEU content (watts) ..................................... 3-53.3 Maximum "quasi steady-state" temperatures during NCT for the ES-3 100 shipping container

with various content heat loads-Kaolite density of 19.4 lb/ft3 and Borobond4 ............. 3-73.4 Maximum "quasi steady-state" temperatures during NCT for the ES-3 100 shipping container

with various content heat loads-Kaolite density of 30 lb/ft3 and Borobond4 .............. 3-83.5 ES-3 100 shipping container maximum steady-state temperatures with Cat 277-4 (100°F

ambient temperature, no insolation) .......................................... .3-9.3:6 ES-3 100 shipping container maximum "quasi steady-state" temperatures during NCT

with various content heat loads and Cat 277-4 (100°F ambient temperature, with insolation) 3-103.7 ES-3 100 shipping container HAC maximum temperatures (Kaolite 1600 density of 19.4 lb/ft3

and Cat 277-4 density of 100 lb/ft") .............................................. 3-123.8 ES-3 100 shipping container HAC maximum temperatures (Kaolite 1600 density of 30 lb/ft3

and Cat 277-4 density of l10 lb/ft3 ) ... : ....................................... 3-133.9 Maximum HAC temperatures recorded on the test packages' interior surfaces ............. 3-143. 0 Total pressure inside the containment vessel at 87.81°C (190.06'F) .................... 3-153.11 Total pressure inside the containment vessel at 123.85 0C (254.930F) ................... 3-163.12 Thermal properties of the materials used in the thermal analysis ....................... 3-173.13 Mechanical properties of the materials used in the static stress analyses ................. 3-193.14 Packaging material technical specifications ........................................ 3-203.15 Component allowable service temperature and pressure .............................. 3-223.16 Summary of results of evaluation for the ES-3 100 under NCT .......................... 3-263.17 Summary of results of evaluation under HAC for the ES-3 100 shipping

arrangement using bounding case parameters ................................... 3-263.18 ES-3 100 test package weights before and after 10 CFR 71.73(c)(4) HAC thermal testing .... 3-343.19 Thermax temperature indicating patches for test units ............................ 3-34

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3.20 Predicted temperatures adjustments (°F) for containment vessel due to HAC ............. 3-363.21 Predicted temperatures of the containment vessel due to HAC ('F) ...................... 3-374.1 Containment Requirements of Transport for Type B Packages' ......................... 4-14.2 Summary of the containment vessel design and fabrication acceptance basis .............. 4-14.3 Isotopic mass and weight percent for the HEU contents ............................... 4-54.4 Activity, A, value, and number of A2 proposed for transport .......................... 4-54.5 Regulatory leakage criteria for NCT ............................................... 4-74.6 Containment vessel verification tests criteria for NCT ................................ 4-74.7 Regulatory leakage criteria for HAC ..... : ........................................ 4-94.8 Containment vessel design verification tests for HAC ............................ ... 4-95.1 Calculated external dose rates for the ES-3 100 package with 36 kg of HEU metal contents

(m rem /h) .......... ........................................................ .. 5-25.2 Calculated external dose rates for the ES-3 100 package with 24 kg of HEU oxide contents

(mrem/h) ............................................................. 5-25.3 Radioisotope specification for all ES-3 100 package analysis source calculations

with HEU content and other nuclides per HEU unit weight .......................... 5-35.4 Photon source for one gram of HEU for all contents ................................. 5-45.5 Neutron source for one gram of HEU for all contents ............................... 5-45.6 Geometric data for the shielding analysis models of the ES-3 100 shipping package as shown

in Fig. 5.1 for NCT ......................................... 5-55.7 Detector locations relative to the drum for NCT and to the. containment vessel for HAC ..... 5-75.8 Shielding model material specifications for the ES-3 100 package with HEU content ........ 5-85.9 ANSI standard photon flux~to-dose-rate conversion factors ........................... 5-125.10 ANSI standard neutron flux-to-dose-rate conversion factors ........................... 5-136.1a. Summary of criticality evaluation for solid HEU metal cylinders and bars ................ 6:56.lb. Summary of criticality evaluation for solid HEU metal slugs ........................... 6-96.1c. Summary of criticality evaluation for solid HEU metal of unspecified geometric shapes

characterized as broken m etal ................................................... 6-126.1d. Summary of criticality evaluation for HEU product and skull oxide .................... 6-196.1e. Summary of criticality evaluation for UNX crystals and unirradiated TRIGA reactor.

fuel elem ents ................................................................ 6-236.2a. HEU fissile material mass loading limits for surface-only modes of transportation ......... 6-286.2b. HEU fissile material mass loading limits for air transport mode of transportation .......... 6-296.3. .Deformation of 18.37-in.-diam ES-3 100 drum projected by finite element analysis

Case "3100 RUN IHL Lower Bound Kaolite May 2004" ............................. 6-476.4. Material compositions used in the ES-3 100 calculation models ........................ 6-537.1 Certified replacement parts for the ES-3 100 packaging ................................ 7-28.1 Acceptance tests for the drum assembly ........................................ 828.2 Acceptance tests for the containment vessel assembly .............................. 8-3

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0

XVI


ABBREVIATIONS AND ACRONYMS

ALARA as low as reasonably achievableAM as-manufacturedANC Average Net CountANSI American National Standards InstituteAS allowable stressASME American Society of Mechanical EngineersASTM American Society for Testing and MaterialsCat 2.77-4 Thermo Electron Corporation' Catalog No. 277-4TM (or Cat. No. 277-4)CD capacity dischargeCERCA Compagnie pour I'Etude et la Realisation de Combustibles AtomiquesCFR Code of Federal RegulationsCMTR certified material test reportCoC Certificate of ComplianceCSI criticality safety indexCV containment vesselCVA containment vessel arrangementDOE U.S. Department of EnergyDOT U.S. Department of TransportationEPDM ethylene-propylene-diene monomerETP explicit triangular packFEA finite element analysisH/X ratio hydrogen-to-fissile isotope ratioHAC Hypothetical Accident ConditionsHEU highly enriched uraniumIAEA International Atomic Energy Agencyk0tt calculated neutron multiplication factorLOD loss on dryingLTL lower tolerance limitM.S. margin of safetyMNOP maximum normal operating pressureMOCFR moisture fraction inside the containment vesselMOIFR moisture fraction of the package external to the containment vesselNCT Normal Conditions of TransportNLF neutron leakage fractionNRC U.S. Nuclear Regulatory CommissionNTRC National Transportation Research CenterOECD Organization for Economic Cooperation and DevelopmentORNL Oak Ridge National LaboratoryPGNAA Prompt Gamma-ray Neutron Activation Analysisppb parts per billionppm parts per millionQA quality assuranceQCPI Quality Certification and ProcurementRCSB Rackable Can Storage BoxSAR safety analysis report

Corporate name changed to Shieldwerx

xvii


SCALE Standardized Computer Analysis for Licensing Evaluationsi standard errorSRS Savannah River SiteSS304 type 304 stainless steelSST/SGT Safe-Secure Trailer/Safeguards TransporterTGA thermogravimetric analysisTI transport indexTID tamper-indicating deviceTS test sampleUNH uranyl nitrate hexahydrateUNX uranyl nitrate crystalsUSL upper subcritical limitVF Volume FractionY-I2 Y-12 National Security Complex

xviii

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REVISION LOG

SARDate Revision Description Affected Pages

No.

38407 0 Original issue All

08/15/05 0, Page Page changes resulting from title page, iv, xxiii, 1-4, 1-145, 2-2,Change I Responses to Request for Additional 2-3, 2-6, 2-31, 2-32, 2-33, 2-34,

Information #1, Y/LF-747. 2-57, 2-59, 2-61, 2-107, 2-125,

2-131, 2-171, 2-173, 2-181, 2-183,2-185, 2-186, 2-189, 2-367, 2-458,2-675, 8-8, 8-9, 8-31

38753 0, Page Page changes resulting fromChange 2 Responses to Request for Additional All Sections

Information #2, Y/LF-76 1.

38795 0, Page Page changes resulting from 1.38, 1.48, Appendix 1.4.1, 2-120,Change 3 Responses to Request for Additional Table 6.4

Information #3, Y/LF-764.

38844 0, Page Added polyethylene bottles and Various pages in chapters 1, 2, 3Change 4 nickel alloy cans as convenience and 4.

containers for authorized HEUcontents. (CoC Revision 1)

08/21/06 0, Page Revised equipment specifications for Appendices 1.4.4 and 1.4.5.Change 5 Kaolite and 277-4 neutron absorber.

(CoC Revision 3)

11/15/06 1 Updated definition of pyrophoric All Sectionsuranium.Evaluated air transport.Revised criticality safety calculationsto remove bias correct factors.Added a CSI option of 3.2.Increased mass of off-gassingmaterial allowed in containmentvessel.Increased carbon concentration inHEU contents.Increased Np-237 concentration isHEU contents.Added uranium zirconium hydrideand uranium carbide as contents(TRIGA fuel).Revised equipment specifications for277-4 neutron absorber. (CoCRevision 3)

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No.

3/29/07 1, Page Updated definition of TRiGA fuel for title pages, viii, xi, xx, 1-12, 1-13,Change I air transport and added TRIGA- 1-20, 6-30, 6-54, 6-64, 6-66, 6-87,

related criticality 'safety cases. 6-119, 6-240 to 6-286, 6-385 to end

5/31/07 1, Page Revised SAR in response to RAIs title pages, xiii, xx, Section 1 andChange 2 dated May 9, 2007 in reference to Section 6

CoC Revision 4

6/30/07 .,',"Page Revised SAR in response to RAls title pages, •able of contents,Chang 3 dated May 9, 2007 in reference to Sectioni 1, and Section 7

CoG Revision 5

Note on revisions: Latest revision is shown as:

0. Additions or changes are indicated by highlighted texto Deletions are indicated by a mark in the margin

0

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Y/LF-717/Rev I/page change 3/Intro/ES-3 00 HEU SAR/sc/06-30-07

•<0.040 lug 2 32U/gU, • 50.0 lug233U/gU with the balance of the uranium being 238U).

Group 2 oxides are in the form of UO,. Material from this group contains at least 20.0% uraniumby weight and displays typical isotopic content (•<0.977 g 235U/g U, •<0.014 g 234U/g U, •0.010 g 236U/g U,•<0.040 lug 232U/g U, <50.0 jIg 233U/g U with the balance of the uranium being 238U).

Group 3 oxides are contaminated with up to 40 jig Pu/g U and are in the form of UO,. Material fromthis group contains at least 83.0% uranium by weight and displays typical isotopic content for uranium(•0.977 g 235U/g U, <0.01 4 g 234U/g U, •0.01 0 g 236U/g U, •0.040 jig 232U/g U, •<50.0 [Ig 233U/g U with thebalance of the uranium being 238U).

Group 4 oxides are in the form of U30 8. Material from this group contains at least 83.0% uraniumby weight and displays typical isotopic content (A0.977 g 235U/g U, •<0.014 g 234 U/g U, -0.010 g 236U/g U,:0.040 jig 232U/g U, •50.0 jig 233U/g U with the balance of the uranium being 238 U).

Group 5 oxides are in the form of UOx. Material from this group contains at least 20.0% uraniumby Weight and displays typical isotopic content (•0.977 g 235U/g U, •<0.014 g 234U/g U, •<0.010 g 236U/g U,•0.040 jig 232U/g U, •<50.0 jig 233U/g U with the balance of the uranium being 238U). This material maycontain considerable activity in the form of unspecified beta emitters.

Group 6 oxides are in the form of UO,. Material from this group contains at least 20.0% uraniumby weight and may display unusually high isotopic concentrations of 233U, 234

U, and 236U (•0.977 g 235U/g U,•0.020 g 2 34U/g U, •0.40 g 236U/g U, •0.040 jig 2 32U/g U, •<200.0 jig 2 33U/g U with the balance of the uraniumbeing 238U).

Group-7 oxides are in thefonnofU 3O8 ..: Material fromi this group is ami..xtur eofgraphiteand U3 0.

The uraniium, conicentration s§,up~to 4.5% by weight ad thecarbon concentration is u~p to 171,000 [ig'gU.Enrichment isupm :to 93.2%,by weight., Concentrationsof other uraniumiisotopes are f;.%014 g -34 U/gjU4

.0 g U/gU 0040g UgU,50.0 g iththba e of the uraium bei .Thsame carbon concentration limit shall apply toany ium carbid content such :,UC, UG2, or U•C3 .

The oxides in Groups 1, 3, 4 and71 are high purity uranium oxide purity (the remainder is only traceimpurities). Oxide Groups 2, 5, and 6 are listed to contain at least 20% uranium by weight, which allows upto 80% non-uranium material. As oxides, depending on the purity and chemical form, 3% to 17% of the totalmaterial composition will be oxygen, leaving up to 77% impurity or "filler". These three oxide groupsinclude a range of scrap and recovered materials. For the least pure uranium oxides, the majority of the fillermaterial is aluminum oxide (from recovered alumina traps or from oxidized uranium-aluminum alloys).Other materials that occur in appreciable quantities in some scrap materials are oxides and compounds ofboron, calcium, iron, sodium, lead, zinc, magnesium, copper, molybdenum, and tungsten. These materialsare essentially inert from the standpoint of criticality safety and chemical interaction with the ES-3100convenience cans and bottles identified in this section for the shipment of oxides.

HEU Metal and Alloy

HEU metal and alloy (alIoysof uranium with aluminum or molybdenum) may be in the form of solidgeometric shapes. Solid shapes may include the following:

1. sphee aeotincidd as a conitentsae

2. cylinders having a diameter no larger than 4.25 in. (maximum of one cylinder per conveniencecan);

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3. square bars having a cross section no larger than 2.29 in. x 2.29 in. (maximum of one bar perconvenience can); and .

4. slugs having dimensions of 1.5 in. diameter x 2 in. tall (maximum of 10 per convenience can).

it-hm the.exceptionof slug content, solid HEU • etl and alloyconten~otofspec •i0dgme apss hail~beahlmited togone item pertconvenience container. HEU bulk metal and alloy contents not covered bythe geometric shapes category specified above will be in the broken metal category, and will be so limited.

HEU bulk metal and alloy contents in the broken metal category may be of unspecified geometricform. HEU bulk metal and alloy in this category may also be of a specific shape where one or more of thecharacteristic dimensions vary from piece to piece (i.e., the height, width, length, radius, etc.).

For pyrophoric considerations, HE•ie and alloy must meet the followigrestritions:

. Uranium metal and alloy (broken) pieces must have a surface-areto-ms rtioof nt trthan I cm 2/g or must not pass through a 3/8-in. miesh sive.

2.Pat...cles and s.al Ishapes w.hich donotpass.the sizerestriction..testsin , and powders, foils,turnings, and wires, are nt permitted unless they are either in a sealed., inerted container or areIIstabilized to an oxide prior to sh ipment (oxidati~on must be complete). If the uranium had beenstored in water, it must be shipped in a sealed, inerted container or completely oi3derior toshipment.t

Metal and alloy may be shipped in tinned-carbon steel, stainless steel, or nickel-alloy conveniencecans.

Uranyl Nitrate Crystals 0Uranyl nitrate crystals (UNX) are formed by dissolving uranium metal or any of the uranium oxides

in nitric acid. Uranyl nitrate hexahydrate (UNH) has a chemical formula of UO,(NO3)2 +6 H, 0. This mostreactive form is used as the bounding composition for uranyl nitrate crystals in the criticality evaluation.Therefore, for UNX contents, X must be less than or equal to 6. The theoretical density of UNH crystals is2.79 g/cm3; however, the working densities will be less.

The user of the ES-3 100 for UNX shipments will be required to use non-metallic containers only (suchas Tefloui or polyethylene bottles) as the convenience container.

TRIGA Fueli Elements

Fuel ,pellets firc-n Training ' sacI~oe and Gene~ral Atomics (TRJGA) reactor elem rentsare authorized to be shipped'in the :ES-3 100. The fuel shall1beuiiirradiated. The TRJGA fuel shall be inthe form of Uranium zirconium "hydride (UZrHx), wherex<Q 2.2Fuel pellets from three types of TRIGA~ifue eements are allowed; TRIGA' Standard Fuel Elements (SEE), Instrumented TRIGA Standard FuelElements (FTC), and TRIGA Fuel Follower Control Rods (FFCR). These fuel elements have threefuelpellets (or sections) per element. The fuel pellets from the SFE's and FTC's to be shipped are 8.5 wt%"Vuranium and 70%o:enriched. Fissile loading is 45.33g g 2 :U per'pellet•(136 g TU per element) and thedimensions are 5 inchesinm. length and 1.44 inches in diameter.The fuel pellets from: the FFCR's•t •6eshipp• d ar.e 8.5,'Nwt% /urapi'urr fand 70% enriched. Fisslloiading :is 37.33 3 g 2 1U per pele.1 1,g: 3W. r

element) and the.diiensionis are 5 inchesin lengthandi131 inches in diameter. SpecificFRIGA fuel data

is given in Tab~el&4T1

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APPENDIX 1.4.10

PYROPHORICITY OF URANIUM METAL

Prepared by:

Reviewed by:

Russell W. SchmidtBWXT Y-12 L.L.C.November 2006Revised June 200

Sanford G. BloomBWXT Y-12 L.L.C.November 2006Re-reviewed June 2007

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APPENDIX 1.4.10

PYROPHORICITY OFURANIUM METAL

I. Introduction

The ES-3 100 is a new shipping container designed for safe and efficient transportation of highly enricheduranium in a wide range of material forms. The ES-3100 has been certified for use with a variety ofcontents, including enriched uranium metal and alloy. However, the current size limitations on uraniumcontents in the form of broken metal are unnecessarily restrictive. The size limits exist because undercertain conditions uranium metal and some uranium alloys are pyrophoric - they have the potential tospontaneously ignite. The size restrictions are intended to eliminate the possibility of spontaneous ignitionduring transport.

The purpose of this analysis is to evaluate the potential for uranium metal pieces to spontaneously igniteunder the conditions expected for shipment in the ES-3 100 shipping container, and to identify limits onthe uranium metal content that will prevent spontaneous ignition during transport, while still allowing ahigh degree of flexibility and utility.

It is important to remember that the content limits developed for the ES-3 100 must be implemented in thefield. Therefore the criteria derived from this evaluation must be simple, robust, and readily applied in all*of the-facilities using this package.

This evaluation does not include new laboratory tests of uranium metal ignition parameters. Such testshave been performed and well documented in the past. This evaluation draws on the extensive body ofexisting data and proven storage and transport practice to identify the bounds within which uranium metaland alloys can be safely transported in the ES-3 100 shipping container.

II. Proposed Definition of Pyrophoricity

The Certificate of Compliance (CoC), gives the definition of broken metal in paragraph 5.(b)(1)(ii), asfollows:

For metal or alloy defined as broken metal, mass limits are specified in Table 2. Uranium metaland alloy pieces must have a surface-area-to-mass ratio of not greater than I cm2/g or must havea mass not less than 50 g, whichever is most restrictive. Powders, foils, turnings, wires, andincidental small particles are not permitted, unless they are restricted to not more than 1 percentby weight of the content per convenience can, and they are either in a sealed, inerted container orare stabilized to an oxide prior to shipment.

It is suggested that thedefinition of broken metal in paragraph 5.(b)(1)(ii) be revised to the following:

For metal and alloj defined as 'boken metal, mass limits are specified i'Table n Uraniummetal and "alloy pieces 'must have a surface-area-to-mass ratio'of not greater than, 1 cm- ormust not pas freely thro•ugh a 3/8-inch (•. O095m) mesh sieve. Particles and small shazpes haidOnot pass this size restriction, as well as powders, foils, -turnings, and-awires, are not permittedunless they are either 'in a sealed container under an inert cover ;gas oriarecopito an oxide prior~to shipment.. .Uranium metal 'or alloy which has been stored in water must beshipped in a sealed container under an inert cover gas or completely oxidedprioro shipment.

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III. Rational for Proposed Changes

The proposed text makes the following changes:

0 The 50 g minimum piece size is eliminated;* A 34/8-in. mesh limit is added;" The phrase "whichever is most restrictive" is deleted;" The 1% limit on inerted material is eliminated.

The I cm 2/g maximum specific surface area limit is the most significant limit in the original text, and thatlimit is retained unchanged. Specific surface area is the most significant parameter in determining if agiven piece of uranium is at risk of spontaneous ignition under a given set of conditions,. and therefore itis appropriate that this restriction should control any other restrictions to the package contents. Thediscussion section below will demonstrate that the I cm2/g maximum allowable specific surface area isadequate to prevent spontaneous ignition in the ES-3 100.

The.50 g minimum piece sze isiioverly restrictive, and•is inconsistent with :t eIcm3g upper'iimit onspecific surface area. Smooth uraniu metal pieces can have a mass of less thlan00g.5gandstill have Aspecific' surface~area lessthan I cm/g (see example 1).lThis makes the 50 g limit to•orders of magnitu , detoo large. The 50 glihmit'is nearly an order of magnitude too large even considerign etal •pipees terough Surface of bro0ke rmetal instead of a 'mooth caasmor poaishedsurfce'

Example 1.donsidera smoottiranirmetal sphere with a'diametier of 0.32 cmi(06 126 in).

,,The ¢adius of the~sphere is0:16 cm'. Thedensity of uranium metal is 19 dgcinThevo'6lume isV -(4/3.)ýPi- r= (4/3)- Pi•(0.163•y'•0.1716CMi "

TheImass is M= Density. V-= 19g/cmn,. 0.016167cm3 =-0:3260 gThe surface arr~eaC .is ,-A =,4: *Pi r2 ; =4 - Pi }0' (O,6Y 0.32 17 cmThe speificýsurfice ar'e~ais SA- A/M-110.3217 cm' /¶0.3260 g =0,9,Wý60~N~/

~This specific -surface.area i's jut hi the I CM2/g Uperij

If instead of a smooth surfacethels'phere has a rough surfýicec iaracterstlc of bro'ken~irnmmmetal, a largersize is needed to ensure that the 1•cm2/glini•mt•lsmaintainmd'l Fo< a 6 roughness

factorof three (meaning that the rough sufdce has an actual surface'areaathst isteet1imestlesurface area&calculatedfro6m the radius), the radius wouldzneed to be three times the aboveexample to giveithe~same specific area. In this case:

r 3*0. 16 cm =0.48 cm; VY (4/3) - Pi (0.48)1~?6 nM•--19g g/cm :. 0•.463 cm 3 = 8.802 g; A = 4. Pi, (0.48)2. rAnd so the:specific surface area' is8.686 cm/8.802g0.9-9868.cm'/gasbabove.

Depending on the surfaceroueghnss;,a sphere with amassbetween 0 33 ,and&8 8 grams willmeet the tc'm2/gspeci~i~c-,surtace ,arkea imiit-.

The I cm 2/g specific surface area limit controls the parameter that is most important in terms .ofpreventing spontaneous ignition, but it is not easy to measure or to use in the field. A mass limit (similarto the 50 g limit in the existing certificate, but more consistent with the I cm 2/g specific area limit) couldbe used, but it is very time-consuming to weigh every piece of metal in a package. An approach that isboth effective at enforcing the 1 cm 2/g specific surface area limit and quick and easy to use in the field isto separate large pieces from small ones in a sieve. The recommended text stipulates a 38-in. mesh sieve

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to quickly remove small particles (with a large specific surface area) from large particles which have asmall specific surface area.

An inbs ate 3/8•i.. s bvghosummgd spheres 3/8 in. (0.95tcm) in diameternmeet the I cmu/ispecific teftuareqsf limit- (a smooth-surfaced osphere of this size has a specific -surface area of 0.33

mlg.Therefore;, a`spher'e which does not pass freely"'through~ the 3/8-i'niimesh 'Sie~ve will, mecttjecii2I seciicsufac aealiit. 'Other simnple shapes such as cubes a~nd~ rods are also effctiel

coritroiled'by the' 3/8in, sieve. Foilsm, tunngs, and ~wires are explicitly forbid'den 'in bothi the, ctireat andproposed text,. unless~they aecnetdt z xd raepca~

The phrase "whichever is most restrictive" has been deleted from the proposed text since the sieve testhas been sized to effectively enforce the I cm 2/g specific surface area limit.

The finalf ciange min the proposedtext 'i to eliminate the 1% of contentweigh m I tlimi on inerted mateal.This limit is unnecessary for uranium metal] sealed in a container containing an in I ertatmospheireoouranium metal that is converted to an oxide. If the uranium metal is converted to an oxide, then it is innodanger of ignition under any conditions, and packing limits for oxides have been explicitly given ithecertificate of comp iance.If the metal has been sealed in a container containing an•inert atmosphere,tqereis no oxygen available to the metal andtherfre no chance of combustion.

Uranium metal packaged for transport in the ES-3 100 is first placed inside a convenience can or othercontainer. These cans are then placed into the ES-3 100 containment vessel (CV). The convenience canswill displace most of the oxygen from the containment vessel, leaving only enough to react with a fewgrams of metal. If a sealed container containing an inert atmosphere somehow came open in transport (anunlikely scenario given the very limited amount of movement possible inside a properly loadedcontainment vessel), this small amount of oxygen is not enough to support spontaneous ignition. Thecontainment vessel has been shown to retain its structural integrity and remain leak tight underhypothetical accident conditions, so no additional oxygen can enter.

IV. Discussion

In his 1995 review,' Terry Totemeier explains pyrophoricity this way: "Pyrophoricity refers to thetendency of certain metals to ignite and burn in a self-sustaining oxidation reaction. The pyrophoricnature of metals is usually defined in terms of an ignition temperature, which is the temperature at whicha metal will ignite and burn in a self-sustained fashion for a given set of conditions." ASTM C-14542

defines. pyrophoric as "capable of igniting spontaneously under temperature, chemical, orphysical/mechanical conditions specific to the storage, handling, or transportation environment".

This evaluation will demonstrate that uranium metal with a specific surface area of I cm 2/g will notspontaneously ignite under the conditions existing in the ES-3 100 during, packaging and transport.

The primary factors determining if the conditions for spontaneous ignition exist are specific surface areaand temperature. Totemeier explains, "Because oxidation is a surface reaction, the amount of areaavailable for reaction is a critical factor in the determination of the heat generated in oxidation. Specificarea is the best parameter to describe the effect of area, as it also accounts for the amount of material notreacting which can serve as a heat sink." Temperature is critical because the amount of heat generated bythe reaction is a function of the reaction rate, which is in turn a function of the temperature. Highertemperatures give higher reaction rates.

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An additional safety factor in the case of the ES-3 100 is the small amount of oxygen available in thesealed inner containment vessel. This serves to limit the total amount of uranium that can oxidize, andtherefore prevents any potential heat build-up from reaching the ignition point of uranium metal.

Ignition Temperature and Transport ConditionsIn figures 4 and 5 of his review, Totemeier plots two separate tests of uranium ignition temperaturesas afunction of specific surface area. For a specific surface area of I cm 2/g these two plots give values of390°C (663 K)a and 340'C (613 K), respectively. Using the lower value and rounding down gives aconservative value of 600 K for the ignition temperature of uranium metal in the ES-3.100.

The ES-3100 thermal analysis determined that the temperature at the containment vessel wall would notexceed 190°F (361 K) for normal conditions of transport (NCT) and 2557F (397 K) for hypotheticalaccident conditions (HAC). These values, particularly the HAC temperature, are very conservative. Theactual results from six separate package tests showed that the CV wall temperature was typically around210 'F (372 K), with the highest recorded value of 241 'F (389 K). Note that all of these temperatures arewell below the 600 K ignition temperature of the uranium metal contents.

Maximum Temperat ure from Oxidation - Basic EquationsUranium metal readily reacts with oxygen to form uranium dioxide (U0 2). This reaction is exothermic.The heat released by the reaction warms the uranium metal, increasing the reaction rate. Under normalconditions for storage and transport, the reaction rate is slow enough that the small amount of heatgenerated by the reaction is lost to the environment, and a stable steady-state is achieved. If the reactionrate is fast enough, and the metal is relatively well insulated, the temperature of the uranium' metal canbuild, slowly at first but at an increasing rate, until the ignition temperature is reached and the metalignites andburns.

The task at hand is to evaluate the balance between heat generation and heat loss in the ES-3 100 underhypothetical accident conditions to verify that a stable steady state is reached, and that the steady-statecondition is safely below the ignition point of uranium metal. A recent paper by Epstein, Malinovic, andPlys3. lays out a useful approach, which will be followed here without the approximations used in theirpaper.

For uranium metal packed in cylindrical cans, the generation of heat throughout the can and the-associated transfer of heat to the can wall is mathematically identical to the generation of heat within awire due to electrical resistance. In their text "Transport Phenomena'' Bird, Stewart & Lightfoot developthe desired relation (equation 9.2-14):

Tcenter - Twall = (S• R2) / (4 . Kth) (Equation 1)

where

a The curve in Totemeierfigure 4 is discontiniuous, with a transition from a lower curve toian upper curveshown at a specific surface area of 6 CM2/g. At a specific surface airea of 1 cnV/g the upper curve wouldgive an ignition temperature of 550 0 C, while extrapolation of the lower curve gives 3900C. The originalreference from which Totemeier drew figure 4 explains that the iranition from the lower curve to theupper curve is inflencedbymany factors, including the metallurgy oftheuranium, any alloygmetas orimpurities, and theý oxygen- conten~t, of t~he gas involved. Therefore this aqalysis useps the 1owe crve valueof 3990)C.

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Tcenter is the temperature at the center of the can (K)Twall is the temperature at the can wall (K)S is the heat production per unit volume (W/cm3)R is the radius of the can (cm)Kth is the thermal conductivity of the uranium contents (W/cm K)

oTeihea• productin per unit volume is af'unctin of the reaction rate and 'th heat> fhe rction. Sincete'oxidation reaction occurs at the surface of ihe uranium, thereaetion rate is typically stated as a massreacted per second per unit area. The specific suirface areath uranium metal density, and the packingdensity are applied to convert the rate per unit uranium surface area to~a rate pr unit ,canv o1!u-meThe~res'ult is:

Rho - * - SAre a*S~RloRate••h2)S~ RhoPhi ~(Equ'ation'.

'Rho•is the denshityofthe uranlim Petal (/cm=)Phi is•the packing densitY'of the uranium in the can:(cm' U+/cy)

<, Skrea is the specific surface area o~f theA urniu (crn'lg)dHxn is the heat of reaction (J/g uranium)RxnRate is ithe reaction rate (g uranium / (s- ,nm 2))

The reaction rate of uranium metal with oxygen has been evaluated by many researchers over the years.The general form of the rate equation used is:

RxnRate KO oP" .. e(T) (Equation 3)

whereKO is the reaction rate coefficient (g uranium / (s . cm 2))P is the partial pressure of water vapor (kPa)n is the exponential coefficient on the partial pressure of water vaporTe is the activation energy (K)T is the temperature, of the reactants (K)

For the case at hand, the temperature of the reactants is the highest at the can center (Tcenter) and thelowest at the can wall (Twall). The average temperature of the reactants is midway between these twovalues. For conservatism the reaction rate of the entire contents is evaluated at the maximum temperature,which occurs at the center of the can (Tcenter). This analysis considers heat transfer only through the canwalls, and ignores heat transfer through the top and bottom of the can. This is also a conservativeassumption.

Equations 1, 2, and 3 are combined to yield:

Tcenter - Twall = (R2- h'o'.,Phi .- iSArea*, drn k"

(Eqpation 4)

Note ,th at Kth is a property of the can contents (that is, for the bed of uranium metal particles with air inbetween), antd not a property of solid uranium metal The thermal 6coniductivity of uranium metal is~ muchhigher than the thermal conductivity of a bed of uranium metal pieces.

P is the partial pressure of HO0 present in the containment vessel at the center-line temperature. Thisvalue is calculated by assuming that the ES-3100 was loaded at ambient conditions of TO and 100%

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relative humidity, which yields a water vapor pressure of P0. When the temperature in the containmentvessel increases the partial pressure of water vapor increases according to the ideal gas law, which in thiscase reduces to:

P = P0 (Tcenter/TO) (Equation 5)

In this evaluation, the maximum possible rate of heat generation due to the oxidation of the ur mmetalis a1so of interest since this heat must be carried avvay by the package, even at HAC. The ma~ximumi rateof heat generatiorfis the heat gene rated by the oxidation of uranium mietal'at the highes eprtrreached during HAC,assummng the maximum allowable load of urpum eSThiis:

Qmax = Mmax - SArea - dHrxn * KO -P". e('Te/Tcenter) (Equation 6)

whereQmax = the maximum rate of heat generation (W)Mmax the maximum uranium metal loading (g)

Evaluation of Thermal StabilityEquation 4 provides the means to evaluate the maximum temperature reached in the uranium metal. Itdoes not by itself validate the stability of the system. This system does present a simple means to evaluatestability - both numerical stability and more importantly physical stability (meaning that the temperaturecannot build to spontaneous ignition).

Solving equation 4 for Twall over a range of Tcenter values and plotting the results produces figure 1:

Figure 1. Thermal Stability of Uranium Metal Package

450 . - . . . .......- -------..___.................. . , --, - / .... 6.00

425 - L..... - - 5.

400.. 4.00

Wall Temperature

EE375 -. 00

Physical stability

d(Tc-Tw)IdTc

350

325 - - - - - -- - - - - - - - -- - - - -- 1 0

Epstei n Stability Parameter

300 0.00

300 325 350 375 400 425 450 475

Center-Line Temp, K .

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The wall temperature calculated from equation 4 and plotted here is the temperature needed in order toprovide enough heat transfer to maintain the given center-line temperature. The wall temperature initiallytracks the center-line temperature - the rate of heat generation is low, so a very small temperaturedifference is sufficient to remove that heat. As the center-line temperature (and therefore the amount ofheat generated) increases, the temperature difference needed to keep the center-line temperature steady atthe given value increases exponentially. At some point (about 435, K in figure 1) the required temperaturedifference gets so large that the wall temperature would have to decrease in order to maintain a stablecenter-line temperature. In an actual package there is no cooling mechanism to do this, and once thecenter-line temperature exceeds this point the amount of heat generated will exceed the ability to carry offthat heat, and the center-line temperature will increase until ejimerthe reaction runs out of oxygen or theignition point is reached. The point at which the required wall temperature stops increasing, marks themaximum stable center-line temperature for the package. The required wall temperature stops increasingat the point where the rate of increase in the needed temperature difference (d(Tcenter-Twall)) equals therate of increase of the center-line temperature (dTcenter). Stated mathematically, when d(Tcenter-Twall)/dTcenter is less than I the required Twall increases along with Tcenter, and stability is maintained.When d(Tcenter-Twall)/dTcenter is greater than I Twall would have to drop to maintain stability asTcenter increases. Since this is not possible in a real package, the temperature in the package wouldsteadily increase to ",either ignitignorn consumptfn'io!ftavailable oxygen. The value of d(Tcenter-Twall)/dTcenter = I marks the maximum stable point for a given package.

Equation 4 is

Tcenter - Twall =(R 2 Rho • Phi * SArea • dHrxn • KO oPn • e(Te/Tcenter)) / (4 Kth)

Taking the first derivative of this equation with respect to Tcenter yields an equation for the stabilityparameter derived above:

d(Tcenter-Twall)/dTcenter = (Tcenter-Twail) °Te/Tcenter2 (Equation 7)

The thermal stability of the system is maintained as long as (Tcenter-Twall) • Te/Tcenter 2 < 1. This.parameter is equivalent to the stability parameter "B" developed in Epstein et al, without the simplifyingassumptions made in that paper. The value of d(Tcenter-Twall)/dTcenter is plotted in Figure 1, along withthe Epstein "B" parameter.

Input ParametersThe key values used to evaluate equations 4, 5, 6, and 7 are shown in Table 1.

Table 1. Ke input arametersParameter Value Units

R 5.27 cmRho 19 g/cm 3

Phi 0.26 cm3/cm 3

SArea I cm 2/gdHrxn 4559 J/g Uranium

KO 76086 gU/(s . cm 2)P0 3.53 kPaTO 300 Kn 0.3

Te 11490 KKth 0.004 W/(cm ° K)

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The sources of these parameters are:

R is the inside radius of a convenience can. The typical can used has an outside diameter of 4.25 in, andan inside diameter of 4.15 in. (10.54 cm).

Rho is the density of uranium metal.

Phi is the packing density of uranium metal 'when packed into th• convenience cans. Opeetor exprience't Y- 2 is that'a.maximum.of 5'kg U.Fbrokeni metal will fit into a " i ODV by 4-875in. high can,which has an internal volume of 14000.6. cmThis uielidsa pack.ing qeigisty of 016 • '. Amii. T• e efie c ofvaiatisnon~6itis \',alue: are disctussed below.

SArea is the specific surface area, which is limited by the package certification to I cm2/g.

dHrxn is the heat of reaction, on a uranium basis. This-value came from Totemeier, page 17 (1089 cal/gU4559 J/gU).

KO, n, and Te are parameters for the reaction rate equation. As noted above, a number of researchers haveanalyzed the reaction rate of uranium metal with various combinations of oxygen and water vapor.Numerous models have been developed from this data. This analysis used the published. model that bestfit the conditions present. Many of the published rate models are only valid up to 100 to 130'C. Thisevaluation requires evaluation beyond 140°C. Many of the models are for either pure oxygen or purewater vapor. McGillivray5 notes that the reaction rate at a given temperature varies with both the oxygenconcentration and the partial pressure of water vapor. The published model that best matches theconditions of this evaluation (reaction in air at temperatures exceeding 140°C, with a small water vaporpartial pressure) is the Pearce model (Pearce, < 100% RH, in Air, T < 192°C) as reported in the Epsteinpaper.

P0 and TO are used to calculate P, the partial pressure of H20 vapor present in the containment vessel atthe center-line temperature. This value is calculated by assuming that the ES-3 100 was loaded at ambientconditions of TO = 300K (80'F) and 100% relative humidity. P0 is therefore the vapor pressure of waterat 300K, which is P0 = 3.53 kPa. When the temperature in the containment vessel increases, the partialpressure of water vapor increases according to equation 5.

Kth is the thermal conductivity of the uranium particle bed. The value here was taken from Epstein et al,page 6. Because of the air-filled void spaces, the bed thermal conductivity is much lower than theuranium metal value of 0.3 W/(cm - K).

ResultsThe set of equations (4, 5, 6, and 7) was evaluated using a commercial software package namedTk!Solver, which has the advantage of being able to automatically iterate to solutions as needed. This isnecessary when solving equation 4 for a fixed wall temperature to -determine Tcenter. Attachment Ishows the rules, input and output for the TK!Solver model for the HAC and NCT cases.

The hypothetical accident condii evaluatin gave a maximum containmentessnel (CV)iatum e(Twall) of255'F (397K). tEvaluatig hequatins 4 through 7 for these conditions yields a Tcenterof 398.4K (257I.5F),,with, a-im'aximum heat ,output (assuming a fuillload of 36 kgU) of 5.9 Watts, iad a stability

jrqaineterTcent•r•-Ta_)dTeter6f ll the th cal2 valueuof•1.0.

0

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The 255'F HAC temperature was basedon auraim hatgeneration of0.4W. The 5.9 W maximumheat generated from the oxidation reaction under those conditions wouldih~eat the CV wall abovethatemperature.•The thermal analysis evaluated the HA for heatgeneratin rates o 0W andO waas, the 0.4 •W standard value. Table 3).7 in. the thermal anal•sis showsthatfor an assumed 20 W heatgeneration inmthe, uraniumnmetal contents, the peak- CVwalltemperature is, 277 'F (409.3 K). Using thisvalue ensures that -the heat transfer from! theuanumcntents e t• e wall andto the rest of thepM;Cge!s;conservati.vely addressed.

Fo'r th 4vsd A all t1emp'eratUre of 277 'F (4093 Keuations 4 through7 ~yield the~ followiniiresults:

dcenter 413. •aK--(284.4 0F) (well below th •eignition tempertur of600 K)MaxQ •289 W-atts, (below the 20W assumption)d(Tcenter-Twall)idTcIenter = 06.272 (well below the critical value of 1.0)

At the NCT wall tempýera4tu-r'e of 1 90'F (360.9 K) the results are> ?Tcenter ' 6.K(9. F) (well beo h giio eprtr of600 K)

M1axQ,=`0.289-Watts (below the 0.4W assumnptioni)4dcTce tr-,Twall)/dTcenter = 0.006 (wellbelow he ritical vaue f 1.0)

These values clearly: show that uranium• metal •uith a spe•ific- suaee-area of no moren than F• Cm/g willnot spoi•naneously !gnite under any anticipated tr•nsport conditions,

Variaion' inInibut -Packing Densit•The -densiwthy " ; '" which " uranim ~metal is packed: Hmto~the6>c inen canstlimitstte sufae6 areat6:available:to' oxidize, Operator experienee at0 -12 is that a maximum •of 5lkgU:Uof:broken :metal will fit

intb:a'4.257in. OD by,4,875-in. high .can, which has an internal• ye! ejof,9OO~cnm:. his yields a padngdensaity•of.0.26 cm3 cm3..

A search ofthe literatiure on, packing densities reveals a lot of'lwork':ok "onsooth spheres, and \very little o0•anything else. Scott and'Kilgour 6 experimented with packin smooth steel:spheres in cylinderis andreported a maximum packing density of 0.6366 after :extensivevibration to compact theesteel spheres asmuchdas possible. The steel balls used in this experiment•iad a.coefficient of friction of 05.2,well belowthe ,value for smooth uranium metal of .0h. Subsequent analysis tby Kong and Lannutti7 considering theeffects o• friction between particies gives packing f•acti6ons ini:the range f0. 0,4j•& 0.46, with higherfictioncoefficients producing lower packing fractions.

The'brok'en metal routinely packed at Y-12 consists oflarge, rough, irregular pieces. The reportedpacking density of 0.26. for this material is consistent with the literature reviewed, particularly;Kong andLannu~tti... Thereforetlhevaite''of 0'.26'Nwýas used 'a's the base value i~n. theanalyse s repo rted above.

To bund the metal contents of the" ES-3 100two additional cases.have been analyzedunder hypotheicalaccident conditions':roughbrokenf metal at a packin Adensi (h e0.46 (the upper thrangereportedby :Kong :& Lannutti); and smooth cast, spheres; ata paielg densi of 064:;(&6i'slgtent with Seqtp4Kilgour),.

For rough 6rokeineal at a pac'king 'density of 0.46 the mriiximur c-'en''r' ieip-erature at HAC was 422 K,well below ~ihe ignitio'n temperature of 600 K. For smooth cast 3/8" ,spheres at a ~packing density of 0.64the maximumriceniter tempe'rature at HAC was 412 K, well below,' theignition temperature of 600 K.

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Oxygen Limitation in the Containment VesselThe analysis above placed no restriction on the amount of oxygen available to react with the uraniummetal contents. In reality, the ES-3 100 containment vessel has a finite volume, which restricts the amountof oxygen available for reaction.

The CV is a cylinder with inside dimensions of 31.00 in. (78.74 cm) tall and 5.06 in. (12.85 cm) indiameter. This produces a volume of 10,215 cm 3, or 10.215 liters. At ambient conditions of 300 K(80.3°F) and 100% relative humidity, 10.215 liters of humid air contains 2.78 g of oxygen from the air,and another 0.23 grams of oxygen in the H20. These 3.01 grams of oxygen can react with 22.40 g ofuranium metal. This is 0.06% of the ES-3100's capacity.

Amass of 24 g of U metal when reacted with 3 .01 g ofoxygen ~willrdieamxiuittlha

Output of 102 W (96.8 BTU), spread out over the time required fothe reacion to take place. This totalamountof o xygen could sustain the NCT maximum heat output of '0.2 Wfor 4.1 days, or theHACpeak heat output for 1.7 hours. If somehow released all at once, the 102 kJ would only raise ~thetemperature of 36 ka of uranium by 24.5 K. More realistically, as< shown by the calculationis aove, anyireaction will..be.low, with enough time for the heat t t ondthe of thepackage.

The ES-3100 CV is 15.1 kg of stainless steel, with a heat capacity of 0.515 J/(g *K). The 102 kJmaximum produced by the oxidation reaction could only raise the temperature of the CV (ignoringcontents) by 13.1 K. This heat sink, plus the heat sinks offered by the CV contents, ensures that theoxidation reaction will not be able to build to the 600 K ignition temperature required for spontaneousignition.

In3 -ratg tho ilIs hn3~f oxygen aviabet ratlwit tiuranium metal inside te~c1osedCV. A il11 16ad of uranimm will, by itself, dsplace nearlirinthe CVThe onvenieecans; spacer :cans ,-and other packing material ill displace even more air, further reducing the amount ofuIraniumnthat, could poss'ibly react A ,the immetal is acked inside closed convenience• cansThese c'ans limii the .oxygen. available to react with -the contensto the oxygen in the convenieniicanitself. Finally, parlt ofthe available oxygen will react with the uranium before the peak HAC conditionsare -reached. Figure22b insection3l shows thatit will take about 4 hours for the CV wall to reach themaximum tempeFrature in the HAC fire. During this four hour temiperature ranmp-up 0.9 grams of oxygenwould be consumed . reaction .with the uranium leaving only 2.1 grams available to react once the HACtemperature was reaclied. Even ifthe uranium surface area was ~uncontrol led, lack of oxygen would ~snuffout any crease in the, uranium reaction rate before it could ro c igition point.

Specific Surface Area Implementation via SieveAs noted above, the I cm 2/g specific surface area limit is not easy to measure or to use in the field. Ascreening method that is simple and easy to use in the field will reduce the potential for packagingmistakes. The recommended approach is to separate large pieces from small ones in a sieve. Therecommended text stipulatesa 3/8-in. mesh sieve to quickly remove small particles (with a large specificsurface area) from large particles which have a small specific surface area.

Exaniple.i Iabove7 showed 'that ie minimum sizei ofa ameatnaspheresmeetiiig' the ,lsemugspecificasurfacearea limit v~aries with the degree of surface 'r6ughness.In' example, lI a sinooth•sphere-1/8 inch in diameterand a rough :sphere: 3/8 inch in diameter both had specific surface areas just below thelI cmr/g limit. The3/8in. :mesh, is stipulated in the recommendedtext in orderto accommodate' both- smooth and roughmetal.

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The actual metal contents of the ES-3! 100 will include both sm h-surfaced and rough-surfaced metal.The smooth-surfaed items include a variety of cast and machined shapes. The ro •o -surfaced items are"broken metal" - large castings that have been fractured into sm~aller pieces. These broken metal pieceswill typically have two or three cast surface wihteremaining 3or 4surfaces of fatured metal.

The 318-inch mesh recommendation is based onii a surface roughness factor of three, mneaning that therough surface has an~ actual~ surface ~area available to react with oxygen that is three timethtoaevlai~ffatr ufcsfrcs~iaii~o shtosmooth-surface of the same gross dimension~s. This rougns factor oftre was drived from an

evalatin offraturesuracesforcasturaiummetal. )Roughness factors ranged from'I.1 to 2.7, withamean value of 2.0. A roughness factor of three was selected to bound thehi•giest value obseNed Th 3/8inch mesh screening is therefore suitable for metal that is fractured on all surfaces. Since as notedaboveeven, broken metal will have several smooth faces the 3/8-inch mesh screening should be quite,conservative.

A&s demonstrated abovea rough-surace sphere whc os not pass freely through the 3/8-in. meshsive will meet th~e I cm2/g specificsuriface, area limit. Other simple shapes such as cubes anid rods arealso effectively controlled by the ,3/8-in. sievre. Foils, turnings, and wires are explicitly forbidden in boththe current and proposed text, unless they ar.converted to'an oxide or are packaged in an inertatmosphere. Attachment 2 shows the dimensions of a variety of shaps that have aspecific surface area ofI em'/g. All of these items will fall through a 3/8-in, sieve, demonstrating that the sieve will effeivelyenforce the I cm2/g specific surface area limit.

Most foils and wires will not fall through a sieve of any reasonable size. The current 50 g test wouldlikewise not reliably exclude these materials, which is why foils, turnings, and wires are explicitlyforbidden in both the current and proposed text, unless they are converted to an oxide or are packaged inan inert atmosphere. Operator training will be required under either the current or the proposed text toensure that these items are properly packaged.

Operator TrainingAs part of the transition to using' ,the new' shipping container, training materials are being prepared Ttoinstruct.the field operations personnel on the proper way to use the ES-3 100. As'-io, above, the trainingfor the operators p acking uranium metalm ito the conveience& cans for shipment in the ES-3100 will be

importantin ensuring that potentiall pyrophoric materials a properly 'categgorized a inerted asnecessary. This training will cover the following points:

. All metal piece -must be evaluated to ensure that their smallest dimension is larger than the 3/8inch'mesh size.

o~ Single solid-metal pieces that are clearly l1arger than the 3/9-inch* mesh in everydimension'do not require sieving

o Items which areobvi4ously unacceptable, such as foils .wiresr'emoved before thesievi Ssng

o Any item that is not' obvi(ously larger than the 3/8-ijich mesh in' ever :dimensioni andlwhich has not been rejected must be sieved.

o Any item that falls through the sieve must'be rejected.0 Operators need, tobe alert to items which may not fall through the sieve but which are too small:

oLong, hin shapes' su~ch as wires and turnings may not fall through tesvewhen shakeni.If the wire or turning could be picked up and poked through the mesh it must be rejected,even if:It'did no •fall through unassisted.

o. ~Wires or turnings may form a tangled ball whichi will not fall through. The abovecriterion applies: if the wire or turning could be•separated and poked through' the mesh itmust be rejected.

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c No distinction is made between wires and rods -~ if the item could be pick~ed up andpoked throughi the mesh it must be rejeced.

o Foils, thin chips or shards - anyitem less than 1/8 inich thick -must be rejected.o Metal showing visible moisture or signs of hav{ing been stored in water must be, reeced~.

Rejected items must be separated for proper hiandlinjg:" Rejected items can be~ shiippedi if packed~ unideran inert cover gas or if converted to an

oxide.o" A aceptable cover gas must be high-purity{>99.99'7%) and dry (S5 ppm moisture).o For metal convertleldto an oxide, Partial oxidation is not acceptable.

V. Conclusion

The evaluations performed show that uranium metal conforming to the I cm2/g limit on specific surfacearea will not spontaneously ignite under any anticipated transport conditions. Spontaneous ignition isindependently prevented by both the I cm2/g limit on the uranium metal and by the limited amount ofoxygen available in the sealed ES-3 100 containment vessel.

The 3/8-in. sieve specified inthe revised text effectively applies the I cm 2/g specific surface area limit tobroken uranium metal in a manner that is quick and easy to use in the field.

0

1-212

Y/LF-7 1 7/Rev 1/page change 3/Ch-1/ES-3100 HEU SAR/sc/06-30-07

Attachment 1 TK!Solver Model and Results

Rules

P=PO*(Tcefitert-O.)Deriv=(Tcenter-TwaII)*Te/,Tcenter'2HRate&SArea*dllrxn*KO* 'P~n~ex(-Te/Tcenter)MaxQ=HRate*MaxMR~t=~e k - P ,*;p-eTetrORate=RRate*32/23 8OTime=-MaxO/Ote

HAC - 277 'F Wall Temperature

InutName.:;

•-,, +Tcenter. ,

409.3 TwallHRateMaxQ

,36000 ~MaxM~3.01 MaxO~

RRate.ORateOTime

Output ~Unit

413.348 K

.000472 W/gU16.9945 W

lU*go

.003728 gU/s

.000501 gO/s1.66822 hr

Comment

Temperature at the centeirof the canTemp at the can wallRate of heat production (w/unlhmiteO 2r )Maximum heat production (from nmaxKgU;)/Maximum ES-3 100 conteits

*Maximumi ES-3 100 Oxygen ContentReaction Rate at given ConditionsOxygen Use rate at given conditi6nsOxygen time to run-out

5.271 ; SArea

Denyv ý'.272221

cmi Can radius*cm A2/gU SpcfcSraeArao atce

.2645597608611490

3100

.004

rhoYIphi

~dHrxn.KOTe~nP0TOPKth

KgL/cmA3 Density of uraniumPacking density, Ucmj CAn 3 mA

J/gU, HKeatof ReatongU/(S*CMA2)~ Recinrt ofiin ere< 6ý

K Recinrt ofiin -Pearce <1 000Reaction rate coefficient -Pear~ce<I 010%

kPa Vapor pressure, of waerat OK Temperature for P0

4.863 73 kPa Vao rsueo Wtra cneW/cm*K~ ThermialConductivity of U particle bed

1-213

Y/LF-717/Rev I/page change 3/Ch-l/ES-3100 HEU SAR/sc/06-30-07

-----------------------------------------------------------------------------

NCT - 190 0F Wall Temperature

Input Name Output

Teenerr 360.969

HRaie1 -03E'6,36000 MaxQ .28899136000 MaxM

3.01 5 MaxO, RRate' . 6.34E-5ORate 80.52E-6OTune 981015

Uniit<

K

W/gUWY5gUgogU/s>~go/i:~hr

Comment

Temperaturte e cener ofttie canTemp atthecan ýyjall1Rate of heat productioni (x{/unliirited 02)jMaximumeat produCtiotn (frrom axmKi j)Maximunm ES3 100 contentsMaximum ES-3 10 OOxygen ContentReaction Rate at given ConditionsOygenUse rate at given' coniithitins,kOxygentime to run-out

5.271 SArea>

Dery, .00607

19.2645.597608611490.33.53300

.004

rhophidHrxn

cm Can radiuscmA2/gLU Specific Surface Ateapof4U particles

d(Tceniter-Twali)/dTcenter

SgO/cm^3 Density of uraniumPackingldensity. U c1 a m1

J/gU Heatof ReactiongU/(s*cmA2) Reaction rate coefficient - Pearce<00%

K. ReactionK rate coefficieint Peairce <It0O%Reation rate coefficient -tPearcej<I00

kPa, V apor pressure of water at TOK- T emeii rature forKPOkPa • apor Pressur ofWatr• r atT. 'T..tW/cm*K, Thermal Conductivity of U le ~be<d

KOTenPOTOPCKth

4.2474

1-214


Attachment 2. Pyrophoric Size Limits on Small Uranium Metal Pieces

Specific Surface Area:P 1.0"'cm'2/g.Uranium Metal Density:,' 19.0 g/cmA3Surface Area Multiplier: 3.0

The surface area multiplier is the ratio of the actual surface area, ,givided, by the simple geometeric surface area.

Spheres - Minimum Safe Diameter .Limiting diameter 0... o9474cm r 0.3730 3in

,Rods - Minimum -Safe DiameterLength Diamiieter,

0.9 ýcm 0.9474 cm r or 0.3730 in10 cm,, 0-9231 cm or 03634 in1.5 cm" 0.8000 cm-6 r 0'3150 in2.0ý cm -0' 7,500~ cm or 0.2953 in2.5 cm 0:7229 cm, or "0.2846 in

infinite , >0.6316 ýcm" or 0,2487 in

Cubes - Minimum Safe Side LengthLimiting side length .0.9474 cm or 0.37.

8.459 grams

1'2.68812.715"14.32616.78819.495

gramsgramsgrams"grams"grams~

Square Cross-Section Rods -Minimum Safe Side LengthLength $Side.1&* 0.9, cm" 0.9474 cm or 0 3730 in

1. 0. cm,', 0.9231 cm" or~ ~0.3634 in" ~ 1.5'cm.. 0.'8000 ~cmn or 0.3150 in.. 2.0 cm 0.7500 cm or 0.2953 in

2.5 'cm 0.7229 cm or 0.2846 inInfinite 0.6316 cm or 02487 in

€

16-155" grams

16.155 grams16. 189 grams18.240 grams21.~375 grams24.822 grams

Chips & Shards - Minimum .Safe ThicknessLength"" Width.' Thickness K

10cm 0. 5 'cm 6.0000,'cmn or<a1'.5 cm 0.5 cm 2.0000 cmo

Scmcm

Infinite

U.: Z0.50,5

1.0 cm1.5 cm2.0 cm2.5 cm3. 0 cm

SInfinite

0.6>0.60.6-

~0.60.60ý6ý

I.1 IIcmcm,cm

Scm

CrM.cm'

,cmcmcm •,cm,

I .JUUU

-1i.'30431.2000:

(;rr,

cmcmcm

or~

,or

'2.3622~0,7874p0.5906~0.51350.47240.3375

2.0000 "cm1.2000o:cm1: 0000 cm0.9091,:". cm0.8571',ccm0.6667 ,cm

orI 0.7874 inor, 0.4724, inor' 0.3937 "in"or :0.3579" inor 0.3375iinor 0.2625, 'in

-or> 0.1243a in>

inininin'inin

57.0O00 grams"28.500 grams28.500 ~grams30.978 grams34.200 grams

22.80020.52022.80025909.

grams'gramsgramsgramsgrams

Infinite. I nfinite : 0.3158i •6cm.'

1-215

Y/LF-717/Rev I/page change 3/Ch- I /ES-3 100 HEU SAR/sc/06-30-07

VI. References

Totemeier, Terry C. ANL/ED/95-2 "A Review of the Corrosion and Pyrophoricity Behavior of Uranium andPlutonium", June 1995, Argonne National Lab-West

2 ASTM C-1454-00, "Standard Guide for Pyrophoricity/Combustibility Testing in Support of Pyrophoricity

Analyses of Metallic Uranium Spent Nuclear Fuel", January 10, 2000, ASTM International

M. Epstein, B. Malinovic, M. G. Plys, SNF-6192-FP, "Uranium Pyrophoricity Phenomena and Prediction", April

2000, Fluor Hanford, Richland Washington

4 R. Byron Bird, Warren E. Stewart, Edwin N. Lightfoot, "Transport Phenomena", 1960, John Wiley & Sons

5 G. W. McGillivray, D. A. Geeson, R. C. Greenwood, "Studies of the Kinetics and Mechanism of the Oxidation ofUranium by Dry and Moist Air - A Model for Determining the Oxidation Rate Over a Wide Range of Temperaturesand Water Vapor Pressures", June 1993, Journal of Nuclear Materials # 208 (1994) pg 81-97.

G. b:. sicott D. MKilgpur "TeDniyo admCoePcig fýhrsFbur 1.969,Brihtis orrial ofApplied PhyscsSrie's 2, Voue2 p 6 6

7C.M. Kong, J.J. Lannutti, "Effect of AgglomerateSizDisributionson Ls99e king,,Fraction", Marchi209,Journal o6f the~ Amer&ican Ceramnic Society', Vl' 3,pp12183 -21"

1-216

Y/LF-717/Rev I/page change 3/Ch-I/ES-3 100 HEU SAR/sc/06-30-07

SECTION 1 REFERENCES

10 CFR 71, Packaging and Transportation ofRadioactive Material, Jan. 1, 2005.

49 CFR Pts. 100-178, Transportation, Oct. 1, 2004.

ASME Boiler and Pressure Vessel Code, An American National Standard, Rules for Construction ofNuclearFacility Components, Sect. 111, Div. 1, Subsection NB, American Society of Mechanical Engineers, New York,2001 ed. with 2002 and 2003 addenda.

ASTM D-2000, Standard Classification Systemfor Rubber Products in Automotive Applications, AmericanSociety for Testing and Materials, Philadelphia, current revision.

Byington, G. A. et al., "Fireproof Impact Limiter Aggregate Package Inside Shipping Container," U.S. patent6,299,950 BI, Oct. 9, 2001.

DG-7003 (Proposed Revision 2 of Regulatory Guide 7.9), Standard Format and Content of Part 71

Applicationsfor Approval ofPackagingfor Radioactive Material, U.S. NRC, December 2003.

MIL-D-6054F, Drum, Metal-Shipping and Storage, June 30, 1989.

NUREG/CR-3854, Fabrication Criteria for Shipping Containers, Lawrence Livermore Natl. Lab., March 1985.

Regulatory Guide 7.10, rev. 1, Establishing Quality Assurance Programsfor Packaging Used in the Transportof Radioactive Material, U.S. N RC, June 1986.

Regulatory Guide 7.11, Fracture Toughness Criteria qf Base Material for Ferritic Steel Shipping CaskContainment Vessels with a Maximum Wall Thickness of 4 Inches (0.1 m), U. S. NRC, June 1991.

NUREG- 1609, Standard Review Plan for Transportation Packages for Radioactive Material, U. S. NRC,March 1999.

1-217

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0

1-218

Y/LF-717/Rev I/Ch-1/ES-3100 HEU SAR/sc/I 1-15-06

Y/LF-717, Rev. , PageChange3Volume 2

SAFETY ANALYSIS REPORT,Y-12 NATIONAL SECURITY COMPLEX,

MODEL ES-3100 PACKAGE WITH BULK HEU CONTENTS

Prepared by theOak Ridge Y-12 National Security Complex

Oak Ridge, Tennessee 378311Managed by

BWXT Y-12. L.L.C.for the

U. S. Department of Energyunder contract DE-AC05-840R21400

Junle 30, 2007

Y/LF-717/Rev 1/page change 3/lntro/ES-3 100 HEU SAR/sc/06-30-07

Y/LF-717/Rev 1/page change 3/lntro/ES-3100 HEU SAR/sc/06-30-07

CONTENTS

LIST O F FIG U R ES .................................................................. xi

LIST O F TA BLES .......................................................... xiii

ACRONYMS AND ABBREVIATIONS ................. .............................. xvii

R EV ISIO N LO G ................................................ ...................... xix

1. GENERAL INFORMATION .............. 1-1

1.1 INTRODUCTION 1....................................................-

1.2 PACKAGE DESCRIPTION .................................................. 1-11.2.1 Packaging ........................................................... 1-41.2.2 Containm ent System .................................................. 1-71.2.3 C ontents ........................................................... 1-91.2.4 O perational Features ................................................. 1-19

1.3 GENERAL REQUIREMENTS FOR ALL PACKAGES ........................... 1-191.3.1 M inim um package size ............................................... 1-191.3.2 Tam per-indicating feature ............................................ 1-20

1.4 A PPEN D IC ES ............................................................. 1-2 11.4.1 CERTIFICATION DRAW ING ........................................ 1-231.4.2 EQUIPMENT SPECIFICATION JS-YMN3-801580-A002,

ES-3100 DRUM ASSEMBLY .......................................... 1-371.4.3 EQUIPMENT SPECIFICATION JS-YMN3-801580-AOO1,

ES-3100 CONTAINMENT VESSEL ............... . .................... 1-471.4.4 EQUIPMENT SPECIFICATION, JS-YMN3-801580-A003,

MANUFA C TURING PROCESS SPECIFICATION FOR CASTINGKAOLITE 1600TM INTO THE ES-3 10 0 SHIPPING PACKAGE ............... 1-61

1.4.5 EQUIPMENT SPECIFICATION, JS-YMN3-801580-A005, CASTINGCATALOG NO. 277-4 NEUTRON ABSORBER FOR THE ES-3100SHIPPING PA CKA GE ........................................ ...... 1-81

1.4.6 PACKAGE CATEGORY DETERMINATION ........................ 1-1211.4.7 HEU OXIDE MATERIAL SPECIFICATION AS PROVIDED BY Y-12

HIGHLY ENRICHED URANIUM DISPOSITION PROGRAM OFFICE ...... 1-1251.4.8 DETAILED ENGINEERING DRAWINGS .............................. 1-1291.4.9 DESIGN ANALYSES AND CALCULATIONS, DAC-PKG-801624-AOO I,

MIXING WEIGHTS AND ELEMENTAL COMPOSITION OF 277-4NEUTRON POISON USED IN THE ES-3100 ............................ 1-171

1.1A0 PYROP•HkR- OFU-R,,NIMME'', AL .................. 1199

SECTION 1 REFERENCES .................................................... 12-7

iii

YILF-717/Rev 1/page change 3/lntro/ES-3 !00 HEU SAR/sc/06-30-07

CONTENTS continued

2. STRUCTURAL EVALUATION .............................. .................... 2-1

2.1 DESCRIPTION OF STRUCTURAL DESIGN .................................... 2-22.1.1 D iscussion ................................................... ...... 2-22.1.2 D esign Criteria ....................................................... 2-42.1.3 W eights and Centers of Gravity ........................................ 2-142.1.4 Identification of Codes and Standards for Package Design ....... ............ 2-14

2.2 M A T ER IA LS ............................................................. 2-182.2.1 M aterial Properties and Specifications ................................... 2-182.2.2 Chemical, Galvanic, or Other Reactions ................................. 2-182.2.3 Effects of Radiation on M aterials ....................................... 2-26

2.3 FABRICATION AND EXAMINATION ........................ ........... 2-262.3.1 Fabrication ................................................... ..... 2-262.3.2 Exam ination ....................................................... 2-28

2.4 LIFTING AND TIE-DOWN STANDARDS FOR ALL PACKAGES ................ 2-312.4.1 L ifting D evices .............. * ......... ............................. 2-312.4.2 Tie-Dow n D evices .................................................. 2-31

2.5 GENERAL CONSIDERATIONS ............................................. 2-322.5.1 Evaluation by Test .................................................. 2-322.5.2 Evaluation by Analysis ............................................... 2-35

2.6 NORMAL CONDITIONS OF TRANSPORT ..................................... 2-352 .6.1 H eat .............................................................. 2-372.6.2 Cold ....................................................... 2-392.6.3 Reduced External Pressure ............................................ 2-402.6.4 Increased External Pressure ........................................... 2-402.6.5 V ibration .......................................................... 2-442.6.6 W ater Spray ........ ........................................... 2-462.6.7 Free D rop ................................................ ......... 2-472.6.8 Corner Drop .................. 2-472.6.9 Com pression ................................................. 2-482.6.10 Penetration .................................................. 2-50

2.7 HYPOTHETICAL ACCIDENT CONDITIONS .................................. 2-51

2.711 Free D rop ......................................................... 2-53

2.7.2 C rush ........................... ................................. 2-642.7.3 Puncture ........................................................ .. 2-702.7.4 T herm al ........................................................... 2-712.7.5 Immersion-Fissile Material ............................ . ..... 2-792.7.6 Immersion-All Packages ........................................ 2-812.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 10' A2) 2-812.7.8 Sum mary of Dam age .................................................. 2-81

2.8 ACCIDENT CONDITIONS FOR AIR TRANSPORT OF PLUTONIUM ............. 2-95

iv


CONTENTS continued

2.9 ACCIDENT CONDITIONS FOR FISSILE MATERIAL PACKAGES FOR AIRTRA N SPO RT ............................................................ 2-95

2.9a SPECIA L FO RM ........................................................... 2-95

2.9b FU EL RO D S .................................................. ........... 2-95

2.10 APPEN DICES ...................................................... 2-972.10.1 ES-3 100 CONTAINMENT VESSEL ASME CODE EVALUATION

(DAC-EA-900000-A006 and DAC-EA-900000-A007) ...................... 2-99

2.10.2 IMPACT ANALYSES OF ES-3 100 DESIGN CONCEPTS USINGBOROBOND AND CAT 277-4 NEUTRON ABSORBERS ............... 2-157

2.10.3 KAOLITE PROPERTIES ................................. ........ 2-4632.10.4 CATALOG 277-4 PROPERTIES ..................................... 2-5352.10.5 BOROBOND4 PROPERTIES ....................... ................ 2-6432.10.6 RECOMMENDED RANDOM VIBRATION AND SHOCK TEST

SPECIFICATIONS FOR CARGO TRANSPORTED ON SST AND SGTTR A ILER S ....................................................... 2-657

2.10.7 TEST REPORT OF THE ES-3 100 PACKAGE, VOLUME 1 - MAIN REPORT,ORNL/NTRC-013/V1, REV. 0, SEPTEMBER 10, 2004 .......... ......... 2-675

2.10.8 TEST REPORT OF THE ES-3 100, VOLUME 3 - APPENDIX K, TU - DATASH E ET S ......................................................... 2-807

2.10.9 PACKAGING MATERIALS OUTGASSING STUDY FINAL REPORT ..... 2-851

SECTION 2 REFERENCES ........................................................ 2-891

3. THERM AL EVALUATION ....................................................... 3-1

3.1 D ISC U SSIO N ............................................................ 3-13.1.1 D esign Features ..................................................... 3-33.1.2 Content's Decay Heat ................................................ 3-43.1.3 Summary Tables of Temperatures ...................................... 3-43.1.4 Summary Tables of Maximum Pressures ................................ 3-11

3.2 SUMMARY OF THERMAL PROPERTIES OF MATERIALS ................... 3-163.2.1 M aterial properties ................................................. 3-163.2.2 Com ponent Specifications ........................................... 3-16

3.3 GENERAL CONSIDERATIONS ............................................ 3-163.3.1 Evaluation by Analysis .................................. ........... 3-163.3.2 Evaluation by Test ...................................... ........... 3-183.3.3 M argins of Safety .................................................. 3-23

3.4 THERMAL EVALUATION UNDER NORMAL CONDITIONS OF TRANSPORT .... 3-273.4'1 H eat and C old .................... .............................. 3-273.4.2 Maximum Normal Operating Pressure .................................. 3-293.4.3 M aximum Thermal Stresses .......................................... 3-29

Y/LF-717/Rev 1/page change 3/Intro/ES-3100 HEU SAR/sc/06-30-07

CONTENTS continued

3.5 HYPOTHETICAL ACCIDENT THERMAL EVALUATION ..................... 3-313.5.1 Initial C onditions ...... ............................................ 3-313.5.2 Fire Test conditions ................................................ 3-313.5.3 Maximum Temperatures and Pressure .................................. 3-333.5.4 Accident Conditions for Fissile Material Packages for Air Transport .......... 3-38

3.6 APPENDICES ......... . ............................................... 3-393.6.1 THERMAL EVALUATION OF THE ES-3 100 SHIPPING CONTAINER

FOR NCT AND HAC (CONCEPTUAL DESIGN WITH BOROBOND4NEUTRON ABSORBER) ........................................... 3-41

3.6.2 THERMAL EVALUATION OF THE ES-3 100 SHIPPING CONTAINERFOR NCT AND HAC (FINAL DESIGN WITH CATALOG 277-4NEUTRON ABSORBER) ........................................... 3-81

3.6.3 THERMAL STRESS EVALUATION OF THE ES-3 100 SHIPPINGCONTAINER DRUM BODY ASSEMBLY FOR NCT (FINAL DESIGNWITH CATALOG 277-4 NEUTRON ABSORBER) ..................... 3-121

3.6.4 CONTAINMENT VESSEL PRESSURE DUE TO NORMAL CONDITIONSOF TRANSPORT FOR THE PROPOSED CONTENTS .................. 3-145

3.6.5 CONTAINMENT VESSEL PRESSURE DUE TO HYPOTHETICALACCIDENT CONDITIONS FOR THE PROPOSED CONTENTS .......... 3-153

3.6.6 SILICONE RUBBER THERMAL PROPERTIES FROMTHERM 1.2 DATABASE ........................................... 3-159

SECTION 3 REFEREN CES ........................................................ 3-161

4. CO N TA IN M EN T ......................................... .................... 4-1

4.1 DESCRIPTION OF THE CONTAINMENT BOUNDARY .................. ..... 4-24.1.1 Containm ent Boundary ............................................... 4-24.1.2 Special Requirements for Plutonium .................................... 4-4

4.2 GENERAL CONSIDERATIONS ............................................. 4-44.2.1 Type A Fissile Packages .............................................. 4-44.2.2 Type B Packages ............................. ...................... 4-4

4.3 CONTAINMENT UNDER NORMAL CONDITIONS OF TRANSPORT(TY PE B PACKAG ES) ...................................................... 4-6

4.4 CONTAINMENT UNDER HYPOTHETICAL ACCIDENT CONDITIONS

(TYPE B PACKAGES) ...................................................... 4-8

4.5 LEAKAGE RATE TESTS FOR TYPE B PACKAGES ........................... 4-10

4.6 A PPEN D IC ES ......................... ................................. 4-114.6.1 DETERMINATION OF A2 FOR THE ES-3100 PACKAGE WITH HEU

CONTENTS ................. 4-134.6.2 CALCULATION OF THE ES-3 100 CONTAINMENT VESSEL'S

REGULATORY REFERENCE AIR LEAKAGE RATES ................... 4-23


vi


CONTENTS continued

5. SHIELDING EVALUATION ...................................... ....... ....... 5-1

5.1 DESCRIPTION OF SHIELDING DESIGN ..................................... 5-15.1.1 D esign Features ..................................................... 5-I5.1.2 Summary Table of Maximum Radiation Levels ........................... 5-1

5.2 SOURCE SPECIFICATION ................................................. 5-1

5.3 DOSE RATE ANALYSIS MODELS .......................................... 5-35.3.1 Packaging Model Conservative Features ................................. 5-75.3.2 Photon model for 36-kg HEU metal content .............................. 5-95.3.3 Neutron model for 36-kg'HEU metal content ............................. 5-115.3.4 Photon model for 24-kg HEU oxide content ................ 5-115.3.5 Neutron model for 24-kg HEU oxide content ............................. 5-11

5.4 SHIELDING EVALUATION ............................................... 5-11

5.5 A PPEN D IC ES ........................... ... ........................... 5-155.5.1 ORIGEN INPUT DATA FROM TABLE 5.3 ............................ 5-175.5.2 CSASN AND ICE INPUT FROM TABLE 5.8 ......................... 5-195.5.3 MORSE ROUTINES AND INPUT DATA ............................... 5-21

SECTION 5 REFEREN CES ......................................................... 5-39

6. CRITICALITY EVALUATION ................................................... 6-1

6.1 DESCRIPTION OF THE CRITICALITY DESIGN ................................ 6-16.1.1 Design Features ..... ........................................... 6-16.1.2 Summary of the Criticality Evaluation ................................... 6-26.1.3 Criticality Safety Index .......................................... I .... 6-4

6.2 PACKAGE CONTENTS ................................................... 6-306.2.1 Fissile M aterial Contents ............................................ 6-306.2.2 Convenience Cans, Teflon and Polyethylene Bottles, and 277-4 Canned Spacers 6-316.2.3 Packing M aterials ............... ................................... 6-316.2.4 Package Content Loading Restrictions .............. .................... 6-32

6.3 GENERAL CONSIDERATIONS ............................................ 6-336.3.1 M odel Configuration ................................................ 6-346.3.2 Material Properties . ........................ 6-506.3.3 Computer Codes and Cross-Section Libraries ............................ 6-616.3.4 Demonstration of Maximum Reactivity ............................... 6-63

6.4 SINGLE PACKAGE EVALUATION ......................................... 6-646.4.1 Solid HEU Metal of Specified Geometric Shapes ........................ 6-656.4.2 HEU Solid Metal of Unspecified Geometric Shapes or HEU Broken Metal ..... 6-686.4.3 H EU O xide ....................................................... 6-68,6.4.4 UN H Crystals ........................ ............................ 6-70

vii


CONTENTS continued

6.5 EVALUATION OF PACKAGE ARRAYS UNDER NORMALCONDITIONS OF TRANSPORT ............................................ 6-716.5.1 Solid HEU Metal of Specified Geometric Shapes ......................... 6-716.5.2 HEU Solid Metal of Unspecified Geometric Shapes or HEU Broken Metal ..... 6-746.5.3 HEU Oxide .................................................... .... 6-756.5.4 UN H Crystals ..................................................... 6-75

6.6 EVALUATION OF PACKAGE ARRAYS UNDER HYPOTHETICALACCIDENT CONDITIONS ................................................ 6-766.6.1 Solid HEU Metal of Specified Geometric Shapes .......... I .............. 6-766.6.2 HEU Solid Metal of Unspecified Geometric Shapes or-HEU Broken Metal ..... 6-776.6.3 HEU Oxide ............................ 6-786.6.4 UN H Crystals ..................................................... 6-78

6.7 FISSILE MATERIAL PACKAGES FOR AIR TRANSPORT ...................... 6-796.7.1 Results for Solid HEU, One Piece per Convenience Can .................... 6-796.7.2 Results for TRIGA Fuel Elements, Three Pieces per Convenience Can ........ 6-836.7.3 Results for HEU Broken Metal, More Than One Piece per Convenience Can ... 6-856.7.4 C onclusions ...................................................... 6-87

6.8 BENCHMARK EXPERIMENTS ........ ................................ 6-876.8.1 Applicability of Benchmark Experiments ............................... 6-876.8.2. Details of Benchmark Calculations .................................... 6-876.8.3 Bias Determination ................................................. 6-88

6.9 A PPEN D IC ES ....................................... ..................... 6-896.9.1 FISSILE CONTENT MODELS ....................................... 6-916.9.2 HAC PACKAGE MODEL ....................................... 6-976.9.3 PACKAGE MATERIAL COMPOSITIONS ............................ 6-1096.9.4 QUALIFICATION OF A NEUTRON ABSORBER MATERIAL

FO R TH E ES-3100 ................................................ 6-1396.9.5 MISCELLANEOUS INFORMATION AND DATA ...................... 6-1516.9.6 ABRIDGED SUMMARY TABLES OF CRITICALITY

CALCULATION RESULTS ........................................ 6-1556.9.7 INPUT LISTINGS OF ES-3 100 CALCULATION MODELS FOR

SELECT CASES IDENTIFIED IN TABLES 6.1a-e ...................... 6-269

SECTION 6 REFERENCES ....................... ................................. 6-445

7. PACKAGE OPERATION S ....................................................... 7-1

7.1 PACKAGE LOADING ..................................................... 7-17.1.1 Preparation for Loading ................... ........................... 7-17.1.2 Loading of Contents ............. ................................ 7-37.1.3 Preparation for Transport ............................................. 7-7

7.2 PACKAGE UNLOADING .................................................. 7-97.2.1 Receipt of Package from Carrier.................................... 7-97.2.2 Rem oval of Contents ................................................ 7-11

viii

Y/LF-717/Rev I/page change 3/Mntro/ES-3100 HEU SAR/sc/06-30-07

r•

CONTENTS continued

7.3 PREPARATION OF EMPTY PACKAGE FOR TRANSPORT ..................... 1-

7.4 OTHER OPERATIONS ........ ....................................... 7-12

SECTION 7 REFERENCES .................................................... 7-13

8. ACCEPTANCE TESTS AND MAINTENANCE PROGRAM .......................... 8-1

8.1 ACCEPTANCE TESTS ..................................................... 8-18.1.1 Visual Inspections and M easurements .................................... 8-38.1.2 Weld Examinations .............................................. 8-48.1.3 Structural and Pressure Tests ........................................... 8-58.1.4 Leakage Tests ....................................................... 8-58.1.5 Component and M aterial Tests .......................................... 8-68.1.6 Shielding Tests ....................................................... 8-68.1.7 Therm al Tests ......................................................... 8-68.1.8 M iscellaneous Tests .............................. ., .................. 8-7

8.2 MAINTENANCE PROGRAM ... ............................................ 8-78.2.1 Structural and Pressure Tests ........................................... 8-78.2.2 Leakage Tests ....................................................... 8-78.2.3 Component and M aterial Tests .......................................... 8-88.2.4 Therm al Tests ................................. ..................... 8-98.2.5 M iscellaneous Tests ............................. .................... 8-9

SECTION 8 REFERENCES ....................... .......................... ......... 8-11

DISTRIBUTION ........................................................... ...... 8-13

lx


LIST OF FIGURES

1.1 Schematic of the ES-3 100 shipping package ......................................... 1-21.2 Exploded view of the ES-3 100 package with bulk HEU contents ....................... 1-31.3 Containment boundary of the ES-3 100 shipping package .............................. 1-81.4 Typical shipping configurations inside the ES-3 100 containment vessel ................. 1-15

1.5 TR GA fuel element ... . . .. . . . .. . . . ... . . . . . . .. . . . . . . 1-202.1 Containment vessel calculated stress locations ..................................... 2-112.2 ES-3100 shipping package center of gravity locations ............................... 2-152.3 ES-3 100 vibration testing arrangement ........................................ 2-442.4 Water spray test arrangement for Test Unit-4 ...................................... 2-452.5 NCT free drop test on Test Unit-4 ........................................... 2-472.6 Compression test on Test Unit-4 ............................................ 2-482.7 Penetration test damage on Test Unit-4 ....................................... 2-492.8 9-mr drop test arrangement for all test units ........... ............................. 2-532.9 9-m drop test damage on Test Unit-4 ....................................... ..... 2-552.10 Cumulative damage from 9-m drop and crush testing on Test Unit-2 .................... 2-572.11 Test Unit-3 damage from 1.2 and 9-m drop tests .................................... 2-582.12 1.2 and 9-m.drop test damage on Test Unit-I ...................................... 2-602.13 1.2 and 9-m drop test damage to Test Unit-5 ....................................... 2-622.14 Cumulative damage following 9-m crush on Test Unit-I ............................. 2-642.15 Cumulative damage following 9-m crush test on Test Unit-3 ........................ 2-662.16 Cumulative damage from 9-m drop and crush testing on Test Unit-4 .................... 2-672.17 Cumulative damage from 9-m drop and crush testing on Test Unit-5 ................... 2-692.18 280 oblique and horizontal puncture tests on Test Unit-I ............................. 2-712.19 40' oblique puncture test on Test Unit-I ........................................... 2-712.20 Horizontal puncture test over Test Unit-I's containment vessel flange ................... 2-722.21 Horizontal CG puncture test on Test Unit-2 ....................................... 2-722.22 24.6' oblique puncture test on Test Unit-3 ........... ............................ 2-732.23 Vertical puncture test on Test Unit-4 ......................................... 2-732.24 Horizontal puncture test over Test Unit-5's containment vessel flange .................. 2-742.25 Visual comparison of the cumulative damage on the crush side surface after the three drop

tests (from top to bottom: Test Unit-I, analytical results with BoroBond, analytical resultswith Cat 277-4) ........................................................ 2-83

2.26 Visual comparison of the cumulative damage on the rigid surface side after the four droptests (from left to right: Test Unit-2, analytical results with BoroBond, analytical resultswith Cat 277-4) ........................................................ 2-86

2.27 Visual comparison of the cumulative damage on the crush plate side after the three droptests (from left to right: Test Unit-2, analytical results with BoroBond, analytical resultsw ith C at 277-4) ........................................................ 2-86

2.28 Visual comparison of the cumulative bottom damage after the three drop tests (from left toright: Test Unit-3, analytical results with BoroBond, analytical results with Cat 277-4) ..... 2-88

2.29 Visual comparison of the cumulative lid damage after the three drop tests (from top tobottom: Test Unit-3, analytical resultswith BoroBond, analytical results with Cat 277-4) ... 2-89

2.30 Visual comparison of the cumulative damage after the three drop tests (from left to right:Test Unit-4, analytical results with BoroBond, analytical results with Cat 277-4) .......... 2-91

2.31 Containment vessel markings at assembly (swivel hoist ring removed prior to testing) ...... 2-932.32 Containment vessel marking after compliance testing ................................ 2-943.1 MSC.Patran axisymmetric finite element model of the ES-3 100 shipping container

with BoroBond4-nodal locations of interest (elements representing air not shownfor clarity). ............................................................ 36

xi

Y/LF-717/Rev 1/page change 3/lntro/ES-3100 HEU SAR/sc/06-30-07

LIST OF FIGURES continued

3.2 Test unit preheat arrangement ........................................... .. 3-243.3 Test unit insertion into furnace .................................................. 3-243.4 Test unit removal from furnace .................................................. 3-253.5 Test unit cool down and monitoring arrangement .................................... 3-255.1 Cylindrical calculational model of the ES-3 100 shipping package for NCT ................ 5-65.2 ES-3 100 HEU metal content radial (top view) geometric models ....................... 5-106.1. R/Z section view of ES-3 100 single-unit packaging model ............................ 6-356.2. R/Z section view at bottom of ES-3 100 single-unit packaging showing

KENO V.a geometry units 1001-1003, and 1006 (partial) ............................ 6-366.3. R/Z section view at center of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1006 (partial) ........................................... 6-376.4. R/Z section view of near top of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1008 (partial) and 1010-1016 ............................. 6-386.5. R/Z section view at the top of the ES-3 100 single-unit packaging showing

KENO V.a geometry units 1016-1019 ............................................. 6-396.6. R/Z section view of ES-3 100 array packaging model ................................ 6-416.7. R/Z section view at the bottom of the ES-3 100 array packaging showing

KENO V.a geometry units 1001-1003, and 1006 (partial) ............................ 6-426.8. R/Z section view at the center of the ES-3 100 array packaging showing

KENO V.a geometry units 1006 (partial), 1007, and 1008 (partial) ..................... 6-436.9. R/Z section view of near top of the ES-3 100 array packaging showing

KENO V.a geometry units 1008 (partial) and 1010-1016 ............................ 6-446.10. R/Z section view at the top of the ES-3 100 array packaging showing

KENO V.a geometry units 1016-1019 ........................ .................. 6-456.11. M odell. ................................................................ 6-496.12. M odel 2 . ..................................................................... 6-496.13. M odel 3 . ................................................................... 6-496.14. Model 4 ............................................................. 6-506.15. Model 5 ........................... .................................. 6-506.16. M odel 6 . ................................................................... 6-506. 16a. 1%j + 2o versus pitch for. trian~gular arrangement of content in CV ............ ~........ 6-66a6.16b. K,, + 2a versus triangular pitch over range of CViimoderation ............ .... 6-66a6.16c. Kj+t2o versus as a function ~of uranium weight fraction and moderation; ......... 6-66b6.17. Kt, vs. enrichment (wt % 231U) for HEU ranging from 7 to 10 kg ..................... 6-816.18. Keft vs. stainless steel shell radius (cm) for 7 kg HEU with enrichments ranging from

19 to 100 "4 % -2 3 U .......................................................... 6-816.19. K, I vs. Kaolite shell radius (cm) for 7 kg HEU with enrichments ranging from

19 to 100 twt % 2 3 U ........................................................... 6-826.20. Keft vs. shell radius (cm) for dry and water-saturated Kaolite for 7 kg HEU at

100% enrichm ent ..... ...................................................... 6-826.21. Kg1f vs shell radius (cm) for 10.4 kg core of UZrH, with stainless steel or Kaolite shell ...... 6-836.22. Ke, vs. core radius (cm) for homogenized core of UZrHx, 500 g polyethylene, and Kaolite

where the Kaolite water content ranges from the dry to the water-saturated condition ...... 6-846.23. Kef vs. excess water from Kaolite for fissile core blanketed with a variable thickness

K aolite shell .................................................................. 6-856.24. K,,,, vs. excess water from Kaolite for core of 1-7 kg enriched HEU broken metal at

100% enrichment, core blanketed with a variable thickness Kaolite shell ................ 6-866.25. K,,, vs. excess water from Kaolite for core 20% enriched HEU broken metal

(0.6-7 kg 235U), core blanketed with a variable thickness Kaolite shell ................... 6-86

Xii


LIST OF TABLES

1.1 U ranium concentration lim its .................................................... 1-101.2 Bounding uranium isotopic concentrations in oxide ................................. 1-101.3 Authorized content and fissile mass loading limits for ground transport ................. 1-13..4 T R GA ,Fuel S e fiatitln s .... ........ ................ . ... ....... .. .. I 2 ý_

2.1 Proposed HE6EU contents for shipment in the ES-3100 ............................. 2-12.2 Category designations for Type B packages ........................................ 2-72.3 Summary of load combinations for normal and hypothetical accident conditions of transport . 2-92.4 Containment vessel allowable stress ............................................. 2-102.5 Allowable stress intensity (S j for the containment boundary construction materials of

construction ................................................................. 2-102.6 ES-3 100 containment boundary design evaluation allowable stress comparisons .......... 2-122.7 ES-3 100 packaging material specifications ........................................ 2-132.8 Packaging weights for various ES-3 100 shipping package arrangements ................ 2-162.9 Com pliance test unit weights ................................................... 2-172.10 Calculated center of gravity for the various ES-3 100 shippingarrangements .............. 2-182.11 Applicable codes and standards for Category I packaging ............................. 2-202.12 Mechanical properties of the metallic components of the drum assembly ................ 2-212.13 Mechanical properties of the lid fastening components for the drum assembly ............ 2-222.14 Mechanical properties of the cast refractory insulation ............................... 2-232.15 Mechanical properties of containment vessel 0-rings ............... .............. 2-232.16 Mechanical properties of the metallic components of the containment boundary ........... 2-242.17 Mechanical properties of the cast neutron absorber .................................. 2-252.18 Summary of NCT- IOCFR71.71 tests for ES-3100 package .......................... 2-332.19 Summary of HAC - I OCFR71.73 tests for ES-3 100 package .......................... 2-332.20 Summary of temperatures and pressures for NCT ................................ 2-362.21 ES-3 100 containment boundary evaluation for both hot and cold conditions .............. 2-412.22 NCT ES-3 100 containment boundary stress compared to the allowable stress

at reduced and increased external pressures ........................................ 2-422.23 Test and analysis summary for the ES-3 100 package ................................ 2-512.24 Recorded height damage to Test Unit-4 from 1.2-m and 9-m drop testing ................ 2-542.25 Recorded diametrical damage to Test Unit-4 from 1.2-m and 9-m drop tests ............ 2-542.26 Recorded diametrical damage to Test Unit-2 from NCT and HAC drop testing .... * ......... 2-562.27 Recorded flat contour damage to Test Unit-2 from NCT and HAC drop testing ........... 2-562.28 Recorded height damage to Test Unit-2 from NCT and HAC drop testing ................ 2-562.29 Recorded height damage to Test Unit-3 from 1.2-m and 9-m drop testing ................ 2-582.30 Recorded diametrical damage to Test Unit-3 from 1.2-m and 9-m drop testing ............ 2-582.31 Recorded height damage to Test Unitýl from 1.2-m and 9-m drop testing ................ 2,592.32 Recorded diametrical damage to Test Unit-I from 1.2-m and 9-m HAC drop testing ....... 2-592.33 Recorded flat contour damage to Test Unit-I from 1.2-m and 9-m drop testing ............ 2-602.34 Recorded height damage to Test Unit-5 from 1.2-m and 9-m drop testing ............... 2-612.35 Recorded diametrical damage to Test Unit-5 from 1.2-m and 9-m HAC drop testing ...... .2-612.36 Recorded flat contour damage to Test Unit-5 from 1.2-m and 9-m drop testing............ 2-622.37 Recorded height damage to Test Unit-1 from the 9-m crush test ........................ 2-632.38 Recorded diametrical damage to Test Unit-I from the 9-m crush test .................... 2-632.39 Recorded flat contour damage to Test Unit-1 from the 9-m crush test ................... 2-642.40 Recorded height damage to Test Unit-3 from the 9-m crush test ......................... 2-652.41 Recorded flat contour damage to Test Unit-3 from the 9-m crush test ................... 2-652.42 Recorded diametrical damage to Test Unit-3 from the 9-m crush test .................... 2-65

Xiii



2.43 Recorded height damage to Test Unit-4 from the 9-m crush test ........................ 2-662.44 Recorded diametrical damage to Test Unit-4 from the 9-m crush test .................... 2-672.45 Recorded height damage to Test Unit-5 from the 9-m crush test ....................... 2-682.46 Recorded diametrical damage to Test Unit-5 from the 9-m crush tests .................. 2-682.47 Recorded flat, contour damage to Test Unit-5 from the 9-m crush test .................. 2-682.48 1--m (40-in.) puncture drop test description and results ............................... 2-702.49 Thermax temperature indicating patches for test units ...................... ....... 2-742.50 Maximum HAC temperatures recorded on the test packages' interior surfaces ............ 2-772.51 HAC ES-3 100 containment boundary stress compared to the allowable stress ............ 2-792.52 Diametrical damage comparison of Test Unit-I with analytical predictions ............... 2-822.53 Flat contour damage comparison of Test Unit-I with analytical results .................. 2-822.54 Cumulative analytical 120 slapdown drop tests maximum effective plastic strain results .... 2-842.55 Diametrical damage comparison of Test Unit-2 with analytical predictions .............. 2-852.56 Flat contour damage comparison of Test Unit-2 with analytical predictions .............. 2-852.57 Cumulative analytical side drop test maximum effective plastic strain results .... ......... 2-872.58 Diametrical damage comparison of Test Unit-3 with analytical predictions .............. 2-882.59 Cumulative analytical corner drop test maximum effective plastic strain results ..... 2-902.60 Diametrical damage comparison of Test Unit-4 with analytical predictions .............. 2-912.61 Cumulative analytical top drop test maximum effective plastic strain results .............. 2-922.62 ES-3 100 test package weights before and after 10 CFR .71.73(c)(4) HAC thermal testing .... 2-923.1 Isotopic mass and weight percent for the HEU contents.............................. 3-43.2 Decay heat for 35.2 kg of HEU content (watts) ...................................... 3-53.3 Maximum "quasi steady-state" temperatures during NCT for the ES-3 100 shipping container

with various content heat loads-Kaolite density of 19.4 lb/ft3 and Borobond4 ............. 3-73.4 Maximum "quasi steady-state" temperatures during NCT for the ES-3 100 shipping container

with various content heat loads-Kaolite density of 30 lb/ft3 and Borobond4 ............... 3-83.5 ES-3100 shipping container, maximum steady-state temperatures with Cat 277-4 (100°F

am bient temperature. no insolation) .............................................. 3-93.6 ES-3 100 shipping container maximum "quasi steady-state" temperatures during NCT

with various content heat loads and Cat 277-4 (100°F ambient temperature, with insolation) 3-103.7 ES-3 100 shipping container HAC maximum temperatures (Kaolite 1600 density of 19.4 lb/ft3

and Cat 277-4 density of 100 lb/ft3) ..................... ........................ 3-123.8 ES-3 100 shipping container HAC maximum temperatures (Kaolite 1600 density of 30 lb/ft3

and Cat 277-4 density of 1 l0 Ib/ftr) ............................................... 3-133.9 Maximum HAC temperatures recorded on the test packages' interior surfaces ............ 3-143.10 Total pressure inside the containment vessel at 87.81°C (190.06 0 F) .................... 3-153.11 Total pressure inside the containment vessel at 123.85°C (254.930F) ................... 3-163.12 Thermal properties of the materials used in the thermal analysis ....................... 3-173.13 Mechanical properties of the materials used in the static stress analyses ................. 3-193.14 Packaging material technical specifications ........................................ 3-203.15 Component allowable service temperature and pressure .............................. 3-223.16 Summary of results of evaluation for the ES-3 100 under NCT ......................... 3-263.17 Summary of results of evaluation under HAC for the ES-3 100 shipping

arrangement using bounding case parameters ...................................... 3-263.18 ES-3100 test package weights before and after 10 CFR 71.73(c)(4) HAC thermal testing .... 3-343.19 Thermax temperature indicating patches for test units ............................... 3-34

xiv

Y/LF-717/Rev I/page change 3/lntro/ES-3 100 HEU SAR/sc/06-30-07


3.20 Predicted temperatures adjustments ('F) for containment. vessel due to HAC ............. 3-363.21 Predicted temperatures of the containment vessel due to HAC ('F) ..................... 3-374.1 Containment Requirements of Transport for Type B Packages ............... ........... 4-14.2 Summary of the containment vessel design and fabrication acceptance basis ............... 4-14.3 Isotopic mass and weight percent for the HEU contents ............................... 4-54.4 Activity, A, value, and number of A, proposed fortransport ............................. 4-54.5 Regulatory leakage criteria for NCT ............................................ 4-74.6 Containment vessel verification tests criteria for NCT ................................ 4-74.7 Regulatory leakage criteria for HAC ......................... ..................... 4-94.8 Containment vessel design verification tests for HAC ................................ 4-95.1 Calculated external dose rates for the ES-3 100 package with 36 kg of HEU metal contents

(m rem /h) ..................................... ............................. ... 5-25.2 Calculated external dose rates for the ES-3100 package with 24 kg of HEU oxide contents

(m rem /h) .................................................................... 5-25.3 Radioisotope specification for all ES-3 100 package analysis source calculations

with HEU content and other nuclides per HEU unit weight ............................ 5-35.4 Photon source for one gram of HEU for all contents .................................. 5-45.5 Neutron source for one gram of HEU for all contents ................... . ......... 5-45.6 Geometric data for the shielding analysis models of the ES-3 100 shipping package as shown

in Fig. 5.1 for N C T ............................................................ 5-55.7 Detector locations relative to the drum for NCT and to the containment vessel for HAC ..... 5-75.8 Shielding model material specifications for the ES-3 100 package with HEU content ........ 5-85.9 ANSI standard photon flux-to-dose-rate conversion factors ......................... 5-125.10 ANSI standard neutron flux-to-dose-rate conversion factors ............................ 5-136.1a. Summary of criticality evaluation for solid HEU metal cylinders and bars ................. 6-56.lb. Summary of criticality evaluation for solid HEU metal slugs ........................... 6-96.1c. Summary of criticality evaluation for solid HEU metal of unspecified geometric shapes

characterized as broken m etal .................................................. 6-126.1d. Summary of criticality evaluation for HEU product and skull oxide ...................... 6-196.1e. Summary of criticality evaluation for UNX crystals and unirradiated TRIGA reactor

fuel elem ents ................ ............................................... 6-236.2a. HEU fissile material mass loading limits for surface-only modes of transportation ......... 6-286.2b. HEU fissile material mass loading limits for air transport mode of transportation .......... 6-296.3. Deformation of 18.3 7-in.-diam ES-3 100 drum projected, by finite element analysis

Case "3100 RUN I HL Lower Bound Kaolite May 2004" . ............................ 6-476.4. Material compositions used in the ES-3 100 calculation models ........................ 6-537.1 Certified replacement parts for the ES-3 100 packaging ................................ 7-28.1 Acceptance tests for the drum assembly ............................................ 8-28.2 Acceptance tests for the containment vessel assembly ................................ 8-3

Xv


0

xvi

Y/LF-717/Rev I/page change 3/Intro/ES-3 100 H EU SAR/sc/06-30-07

ABBREVIATIONS AND ACRONYMS

ALARA as low as reasonably achievableAM as-manufacturedANC Average Net CountANSI American National Standards InstituteAS allowable stressASME American Society of Mechanical EngineersASTM American Society for Testing and MaterialsCat 277-4 Thermo Electron Corporation' Catalog No. 277-4TM (or Cat. No. 277-4)CD capacity dischargeCERCA Compagnie pour I'Etude et la Realisation de Combustibles AtomiquesCFR Code of Federal RegulationsCMTR certified material test reportCoC Certificate of ComplianceCSI criticality safety indexCV containment vesselCVA containment vessel arrangementDOE U.S. Department of EnergyDOT U.S. Department of TransportationEPDM ethylene-propylene-diene monomerETP explicit triangular packFEA finite element analysisH/X ratio hydrogen-to-fissile isotope ratioHAC Hypothetical Accident ConditionsHEU highly enriched uraniumIAEA International Atomic Energy Agencyketi calculated neutron multiplication factorLOD loss on dryingLTL lower tolerance limitM.S. margin of safetyMNOP maximum normal operating pressureMOCFR moisture fraction inside the containment vesselMOIFR moisture fraction of the package external to the containment vesselNCT Normal Conditions of TransportNLF neutron leakage fractionNRC U.S. Nuclear Regulatory CommissionNTRC National Transportation Research CenterOECD Organization for Economic Cooperation and DevelopmentORNL Oak Ridge National Laboratory.PGNAA Prompt Gamma-ray Neutron Activation Analysisppb parts per billionppm parts per millionQA quality assuranceQCPI Quality Certification and ProcurementRCSB Rackable Can Storage BoxSAR safety analysis report

Corporate name changed to Shieldwerx

xvii

Y/LF-717/Rev I/page change 3/lntro/ES-3100 HE U SAR/sc/06-30-07

SCALE Standardized Computer Analysis for Licensing Evaluations standard errorSRS Savannah River SiteSS304 type 304 stainless steelSST/SGT Safe-Secure Trailer/Safeguards TransporterTGA thermogravimetric analysisTI transport indexTID tamper-indicating deviceTS test sampleUNH uranyl nitrate hexahydrateUNX uranyl nitrate crystalsUSL upper subcritical limitVF Volume FractionY-12 Y- 12 National Security Complex

xviii

Y/LF-717/Rev 1/page change 3/Intro/ES-3100 HE U SAR/sc/06-30-07

REVISION LOG


No.

38407 0 Original issue All

08/15/05 0, Page Page changes resulting from title page, iv, xxiii, 1-4, 1-145, 2-2,Change I Responses to Request for Additional 2-3, 2-6, 2-31, 2-32, 2-33, 2-34,

Information #1, Y/LF-747. 2-57, 2-59, 2-61, 2-107, 2-125,2-131, 2-171, 2-173, 2-181, 2-183,2-185, 2-186, 2-189, 2-367, 2-458,2-675, 8-8, 8-9, 8-31

38753 0, Page Page changes resulting fromChange 2 Responses to Request for Additional All Sections


38795 0, Page Page changes resulting from 1.38, 1.48, Appendix 1.4.1, 2-120,Change 3 Responses to Request for Additional Table 6.4


38844 0, Page Added polyethylene bottles and Various pages in chapters 1, 2, 3Change 4 nickel alloy cans as convenience and 4.

containers for authorized HEUcontents. (CoC Revision 1)

08/21/06 0, Page Revised equipment specifications for Appendices 1.4.4 and 1.4.5.Change 5 Kaolite and 277-4 neutron absorber.

(CoC Revision 3)

11/15/06 1 Updated definition of pyrophoric All Sectionsuranium.Evaluated air transport.Revised criticality safety calculationsto remove bias correct factors.Added a CSI option of 3.2.Increased mass of off-gassingmaterial allowed in containmentvessel.Increased carbon concentration inHEU contents.Increased Np-237 concentration isHEU contents.Added uranium zirconium hydrideand uranium carbide as contents(TRIGA fuel).Revised equipment specifications for277-4 neutron absorber. (CoCRevision 3)

XiX

Y/UF-717/Rev 1/page change 3/Intro/ES-3 100 HEU SAR/sc/06-30-07


-No.

3/29/07 1, Page Updated definition of TRIGA fuel for title pages, viii, xi, xx, 1-12, 1-13,Change I air transport and added TRIGA- 1-20, 6-30, 6-54, 6-64, 6-66, 6-87,

related criticality safety cases. 6-119, 6-240 to 6-286, 6-385 to end

5/31/07 1, Page Revised SAR in response to RAIs title pages, xiii, xx, Section 1 andChange 2 dated May 9, 2007 in reference to Section 6

CoC Revision 46130/07 1, Page Revised SARii'*esponse 11-RI s titi estblofcnet

Change.3 datediMay 9, ..200 ..... ..refe..nc. to Se ........ of:cont7tIC oC Revis-io~n-5

0

Note on revisions: Latest revision is shown as:

o Additions or changes are indicated by highlighted text0 Deletions are indicated by a mark in the margin

.0

0

XX


0

Table 6.9.6-18b. HAC results for skull oxide (SO) content in packaging calculation model

Sat.case name SO U 3 0 8 H 2 0 U 2 3 5

U C mg C/ CV H 20 Unident.(a_) (M_ (q) SO h/x WL) () (2) L 235U (Q) CV h/x (g) a k.+20

ciask 9 7 21300 19865 760 2.36 16816 15673 921 58764 647 3.44 1 0.48912 0.00101 0.49114

ciask_9 8 21300 19865 1140 3.00 16816 15673 921 58764 971 4.61 1 0.52104 0.00109 0.52322

hciask 9 9 21300 19865 1521 3.63 16816 15673 921 58764 1294 5.79 1 0.55643 0.00104 0.55851

hciask 9 10 21300 19865 1901 4.26 16816 15673 921 58764 1618 6.96 1 0.59193 0.00105 0.59403

hciask 9_11 21300 19865 2281 4.90 16816 15673 921 58764 1941 8.13 1 0.62726 0.00109 0.62945

hciask_9 12 21300 19865 2661 5.53 16816 15673 921 58764 2265 9.30 1 0.65952 0.00106 0.66163

hciask_913 21300 19865 3041 6.16 16816 15673 921 58764 2588 10.47 1 0.69094 0.00127 0.69348

iciask_9_15 21300 19865 3801 7.43 16816 15673 921 58764 3235 12.82 1 0.75757 0.00116 0.75989

hciask 10 1 21300 20786 0 1.05 17596 16399 0 0 0 1.05 1 0.43087 0.00111 0.43308

hciask 10 6 21300 20786 391 1.67 17596 16399 0 0 324 2.19 1 0.46066 0.00092 0.46250

ciask_10 7 21300 20786 781 2.29 17596 16399 0 0 647 3.32 1 0.49381 0.00097 0.49574

hciask 10 8 21300 20786. 1172 2.91 17596 16399 .0 0 971 4.46 1 0.52679 0.00113 0.52906

hciask 10 9 21300 20786 1562 3.54 17596 16399 0 0 1294 5.60 1 0.56200 0.00121 0.56442

hciask_10 10 21300 20786 1953 4.16 17596 16399 0 0 1618 6.73 1 0.59673 0.00118 0.59908

ciask 10_11 21300 20786 2343 4.78 17596 16399 0 0 1941 7.87 1 0.63042 0.00109 0.63260

hciask_10 12 21300 20786 2734 5.40 17596 16399 0 0 2265 9.00 1 0.66502 0.00116 0.66735

hciask 10 13 21300 20786 3124 6.02 17596 16399 0 0 2588 10.14 1 0.69780 0.00108 0.69997

hciask_10 14 21300 20786 3515 6.64 17596 16399 0 0 2912 11.28 1 0.73223 0.00123 0.73470

hciask 10 15 21300 20786 3905 7.27 17596 16399 0 0 3235 12.41 1 0.76462 0.00131 0.76725

nciask_10 14 21300 20786 3515 6.64 17596 16399 0 0 2912 11.28 1 0.72808 0.00150 0.73109

nciask 10 15 21300 20786 3905 7.27 17596 16399 0 0 3235 12.41 1 0.76141 0.00140 0.76422

6-239

Y/LF-717/Rev l/Ch-6/ES-3100 HEU SAR/rlw/I 1- 15-06

7. PACKAGE OPERATIONS

The ES-3 100 shipping package shall be operated in accordance with applicable Nuclear RegulatoryCommission (NRC), U.S. Department of Transportation (DOT), and other federal, state, and local regulationsto protect the health and safety of the public, workers, and the environment. Furthermore, the ES-3 100 shallbe operated according to a site-approved quality assurance plan.

Specific criteria for operating the ES-3 100 package with highly enriched uranium (HEU) contentsare presented in this section. The packaging user shall develop detailed site-specific operating proceduresbased on these criteria and on the NRC-issued Certificate of Compliance (CoC). These procedures shall bein accordance with 10 CFR 71, Subparts A, G, and H. The package operations should be consistent withmaintaining occupational radiation exposures as low as reasonably achievable (ALARA) as required by10 CFR 20.1101, "Standards for Protection Against Radiation."

7.1 PACKAGE LOADING

The user of the packaging shall:

1. have authorization to acquire, package, transport, or transfer radioactive, fissile, or special nuclearmaterial:

2. have the latest NRC CoC and referenced SAR sections for the ES-3 100 package with HEU contents;

3. comply with all actions and restrictions specified in the CoC;

4. be registered as a user of the packaging with the NRC; and

5. have an site-approved quality assurance program that meets the requirements of 10 CFR 71,Subpart H.

7.1.1 Preparation for Loading

The ES-3 100 Containment Vessel (CV) may be loaded while inside or outside the drum. Thisdecision is a site dependent. Detailed, written operating procedures shall include, at a minimum, the processsteps listed below before the contents are placed in the ES-3 100 package. These steps, initiated by theoperating personnel and their supervisor, ensure that:

1. all appropriate documents have been reviewed by operating personnel and are available for furtherreview, if necessary.

2. the radioactive material contents are authorized by the CoC, and the use of the package complieswith all conditions in the CoC.

3. the packaging has been properly maintained and is in unimpaired condition. (All requiredrefurbishment and periodic maintenance shall have been performed and documented within thescheduled requirements of the CoC, the SAR, and the maintenance program.)

4. A valid leak-test sticker must be present on the containment vessel to ensure that the required

7-1

Y/LF-717/Rev I/page change 3/Ch-7/ES-3100 HEU SAR/sc,106-30-07

acceptance leak test or the annual leak test has been performed.

5. packaging interior, nonfixed surface contamination levels are not high enough to significantlycontaminate the contents. Nonfixed surface contamination limit requirements are given in10 CFR 20.1906, 10 CFR 71.87(i), and 49 CFR 173.443 for alpha, beta, and gamma-emittingradionuclides.

6. all closure fasteners are those furnished with the packaging or are certified replacements and areacceptable for use.

7. all required parts of the packaging and all necessary equipment are available and ready for use.

8. the silicone rubber pads,iif required, have been inspected prior to use.

The user may replace certain parts during loading. Parts that may be replaced by the user areidentified in Table 7.1. The certification of all replacement parts must be traceable. The user must documentthe replacement.

Table 7.1. Replacement parts for the ES-3100 packaging

Part Description Material Specification/Drawinga

Containment vessel 5.359-in. inner diam (ID) by 0.139-in. Ethylene propylene ASTM D-2000inner O-ring diam stock M2E801580A013

Containment vessel 5.859-in. ID by 0.139-in. diam stock Ethylene propylene ASTM D-2000outer O-ring M2E801580A013.

Drum lid washer 0.844-in. ID by 1.375-in. outer diam Stainless steel ASTM A240 or ASTM A276(OD) :K 0.25-in.-thick M2E801580A005

Drum lid hex nut %-in.-I I unified coarse thread (UNC) Silicon bronze ASTM F467 per ANSI B 18.2.2M2E801580A005

Plug (Plastic plug around circumference of Nylon 6/6 62MP0312drum assembly and top of top plug) Micro Plastic. Inc.

M2E801580A002M2E801580A008

Modified VCO Leak-test port plug Brass P/N 04-2126Threaded Plug M2E801580A011

Silicone rubber pads Silicone rubber M2E801580-A009-122 ± 5 Shore A M2E801580-A009-2

M2E801580-A009-3

a. Drawings are available in Appendix 1.4.8.

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Y/LF-717/kev I/page change 3/Ch-7/ES-3100 HEU SAR/sc/06.-30-07

7.1.2 Loading of Contents

The operating procedures for the ES-3 100 package with HEU contents shall be specific regardinghandling of all package components. Approved procedures shall clearly state all safety aspects or activities

such as personnel protection (radiation, chemical, physical); surface contamination or radiation surveys;

nuclear criticality safety; and environment temperature.

The detailed operating procedures for inserting the content into the packaging shall include, at a

minimum, the process steps listed below:

I. the appropriate CV is positioned for packing and verify that the vessel was loaded according to these

steps.

2. the HEU material has been verified as being within the limits specified in this SAR and the NRC

CoC for material mass, material dimension, uranium content, and 235 U enrichment as required inSect. 1.2.3. The content shall be verified using accountability records and weight measurements.

3. all contents and their associated %a.,bottles and packing material are weighed and are within the

allowable weights specified in Sect. 1.2.3.6.

4. the HEU material and associated packing material (convenience cans, spacers, bagging, pads, etc.)

have been inserted as required by Sect. 1.2.3.

5.for proper handlinig 'olbroken metal contentis, pyrophor ity co%-ncerns miust firs~t be addressed.Operatorshav• etwo options: 1) treat all broken metal as pyrophori(in thisase skip to proceurein Itemn #7 below) or, 2) determine whethe~r or n~ot the broken metal is pyrophoric (in this c~ase oto,procedutre in Item ft6 below).~

for pyrophoric categorization of boken metal, -first evaluate the spcific surface area of the metalpieces (surface-area-to mass ratio). If that value isless •hian or equal to 1.00 cm2/g then theimetalpieces are not pyrophoric (and no further action is nieeded). If the specific surfac~e area is greateFthan 1.00 emn2/g, or if the specific saarea can not be evaluated, perform th• remainer of thiSprocedure:

a Metal pieces mustbe physically evaluated to ensure that their, smallest dimension is larger thana 3/8 inch mesh size.

i. Single solid-mnetil pieces that are clearly largerth~an the /8-Iinh mesh in every dimensionidonotf require sieving, and are acceptabli.

ii'. Itemis which are unacceptabl'e by definition(page 1- 1i2), such as powders, foils, wires, andtrn•ngs, should be rejected before the i:eving

iiýi• A~ny itemsthat arenot obV10usly largerthanthe 3,8-inch mesh in every dimension aid

which have not been rejected rmust •be sieved.

iv. Any item that ,falls thi-rot1thhe gsieve must be rejected.

V1. Any rejected item must be 'handled according to the procedure in Item #7belopw.

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Y/I."-7 17/Reo I/page change 3/C0-7/1-S-3 100 lIEU SAR/sc/06-30-07

b., Operators need to be ale~rt*to itemns which mav not fall through the. sieye but still may~bepyrophoric:.

i Longw thin shapes such as wires anrdturnings mayynot fall through the sieve when shakegn.Ifjth'ewire or turning could be picked up and poked through the nmesh it must b•, rjected,even if it did not fall through unassisted.

1... Wiresturnrings, and other items may form a tangled mass which ill not fall through thlesieve. The abovecriterion applies: if th~ewire, l•mng, o other item could be separatedand poked through the mesh it mnust be reeted (unless the specific, Surface arcaof the itemcan be evaluate~d and it is less than or~ equal to I cin2 /g).

iii. Foils, thin chips or shards - any item less than 1/8 inch3 thick - must be rejected.

iv. Metal shown visible moisture or sians of having~ bet tred in water must be rjected.

v. Any rej•cted metal must be handled according to the procedure iin Item .. 7 below.

7 . Rejected items must be separated for proper han~dling:

a. Rejected items can be shipped iin the ES-3 100 if packed in a sealed contaierp under an inert

covjer g~as or if converted to an oxide.

b. Ail ace~ptable cover gas is high-purity argon (Žý99.997%) and dry (<_5 ppri moisture).

c. For metal converted o an oxide, only complete oxidation is cceptable.,

7.1.2.1 CV Assembly and Leak Testing

The detailed operating procedures shall describe activities to prepare the packaging for finalclosure and shipment. They shall include, at a minimum, the process steps listed below when preparingthe containment vessels for closure:

the containment vessel 0-ring grooves and sealing, surfaces are visually checked for scratchesthat may have occurred during insertion. If scratches are found, .Sect. 8.2.2 should be reviewedfor criteria for evaluating surface scratches, possible repair methods for minor scratches, andrejection criteria for significant scratches.

2. the 0-rings and the containment vessel sealing surfaces are free from debris and have not beendamaged during loading operations. lsopropyl alcohol and lint-free cotton cloth or swabs shouldbe used to clean the grooves and sealing surfacds. The 0-rings may be wiped with lint-freegloves, cloth, or swabs. Note that the 0-rings shall be lubricated with a thin coat of Super 0-Lube.

3. the containment vessel sealing lid is secured to the containment vessel body by the containmentvessel closure nut.

4. the closure nut is tightened to 162.7 ± 6.78 N'm (120 ± 5 ft-lb) of torque as specified inDrawing M2E801580A0 11 (Appendix 1.4.8). No impact wrench shall be used.

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Y/I I.-717/Roy I/ p age cIiangc 3/0 i-7/1 S-3 100 FIU SAR/sc/06-30-07

5. the assembled and loaded containment vessel is prepared for leak testing.

6. the annulus between the O-rings shall be leak tested after the CV is loaded in accordance withANSI N 14.5-1997, Sect. 7.6, The user will perform the following steps:

ensure that equipment used to perform leak tests has been properly calibrated to ensure theaccuracy of test measurements

ensure that all leak testing is performed in accordance with a quality assurance program,procedures, and documents the results

* ensure that the test method used has a sensitivity of at least I x 10-4 ref-cm 3/s air

* use either the gas-pressure drop (ANSI N 14.5-1997, Section A.5. l) or the gas-pressure rise(ANSI N 14.5-1997, Section A.5. 1) tests

* remove the modified VCO threaded brass plug from the leak-test port opening

* hook-up the leak test equipment into the leak-test port opening using an appropriate fitting

* either pressurize or evacuate the annulus between the O-rings to a suitable pressure andmeasure the change in pressure and temperature within the test volume during a specifiedtime period

pressure measurements must be accurate within 1% or less and tests should be carried outinisothermal conditions.

calculate the total leakage rate, using the known test volume and test results and ensure thatthe measured leak rate is less than I x 104 ref-cm 3/s air

7. the vacuum coupling is removed.

8. the modified VCO threaded brass plug is tightened into the leak-test port opening.

The user must ensure that their procedure meets the requirements of ANSI N14.5-1997.

If the inner O-ring requires replacement, the containment vessel must be retested per Sect. 8.2.2prior to use. This requirement does not apply to the outer O-ring, as it is not part of the containmentboundary.

Following a successful leak test, the containment vessel with its content is ready to be loadedinto the drum assembly.

7.1.2.2 Drum Closing

A radiation check of the contents may be conducted prior to loading to measure the content doserate, The measured dose rate should be compared with known values for such a test. After loading iscomplete, radiation measurements shall be taken to determine the package dose rate, which establishesthe transport index (TI).

The detailed operating procedures shall include, at a minimum, the process steps listed below

when preparing the drum assembly for closing and sealing. The operating personnel and their supervisorshall ensure that:

7-5


1. the CV assembly with content is ready for loading into the drum assembly or lowering into thedrum assembly, depending on whether it was loaded outside the drum or inside the drum.

2. the drum assembly (with top plug removed) is ready to receive the containment vessel assemblyand that the containment vessel assembly, silicone rubber pads, drum lid, drum-lid nuts andwashers, and tamper-indicating devices (TIDs) are available.

3. the approved lifting equipment is available and in place. For lifting equipment restrictions, seeSect. 7.1.3.1.

4. the CV swivel hoist ring is removed after the CV is positioned in the drum.

5. the CV flange pad is placed on top of the containment vessel and the plug pad is placed on theinner liner shelf.

6. the top plug is placed into position over the CV using eye bolts attached to the threaded holesprovided on the top plug.

7. the eye bolts are removed from the top plug.

8. the drum lid, the drum washers and bronze drum nuts are installed.

9. the nuts are tightened to 40.67 + 6.78 N-m (30 + 5 ft-lb) of torque with no sequence specified. Noimpact wrench shall be used.

10. the TIDs areattached through both TID lugs.

11. the gross package weight does not exceed 190.5 kg (420 Ib).

12. surveys for nonfixed surface contamination and radiation dose rate measurements are conducted.The nonfixed surface contamination survey shall be conducted in accordance with the user'sfacility procedures. The survey shall use criteria that are derived from the surface radioactivityguidance of 10 CFR 20.1906, 10 CFR 71.87(i), or the user's site-specific criteria, whichever isthe most stringent.

13. nonfixed surface contamination is removed as applicable.

14. all "empty" or inappropriate labels or tags are removed from the exterior surface of the package.

15. the package is labeled with the appropriate material description, nuclides, activity/mass, and TIin accordance with 49 CFR 172.403.

16. the package is marked with the minimum marking "Radioactive Material, Type B(U), Fissile,UN3328" in accordance with 49 CFR 172.310.

17. the package radiation dose rate at the surface is measured. The package radiation dose rate atI m from the surface shall be measured to establish the TI for the package and to ensure thatcontent does not exceed the expected or allowable dose rates (see Sect. 5). The analysispresented in the containment evaluation (Sect. 4) has determined that this is a Type B, fissilematerial package.

7-6


7.1.3 Preparation for Transport

7.1.3.1 Package Handling

Criticality Safety Index (CSI) values for the ES-3 100 package with various payloads can befound in Table 1.3.

The ES-3 100 is handled using industry-standard drum-handling equipment. Operatingprocedures shall include requirements to limit clamping pressures on forklift drum-handling equipment toprevent damage to the ES-3 100 drum body (see Sect. 1.2.1.1 for limits on forklift gripping forces).

7.1.3.2 Decontamination

The package may be placed onto areas that are covered by disposable covering, such as plastic orpaper, to reduce the nonfixed surface contamination of physical structures.

The package must be shipped in an enclosed conveyance. Generally, the exterior surfaces of thepackage will remain relatively clean. However, each user shall prepare procedures to clean dirtypackages. These procedures shall, at a minimum, consider the following:

1. The drum is austenitic stainless steel.

2. The drum nut is silicon bronze.

3. The drum vent holes are covered with plastic push-in plugs.

4. The labels and markings on the drum must remain legible.

5. The cleaning solution must be checked for contamination.

7.1.3.3 Requirements Prior to Shipment

The shipper shall ensure that the quality control requirements of 49 CFR 173.475 and the routinedetermination requirements of 10 CFR 71.87 have been satisfied prior to each shipment. Detailedoperating procedures [10 CFR 71.87(f)], shall provide evidence that these requirements are met andensure that:

1. the package is proper for the content shipped and verified with the appropriate records by theuser prior to content loading [10 CFR 71.87(a)];

2. the package is in unimpaired physical condition [10 CFR 71.87(b)];

3. the closure devices of the package are properly installed, secured, and free of defects[10 CFR 71.87(c)];

.4. the containment vessel has been loaded properly and preparation for shipment has been followed,witnessed, and checked;

5. the internal pressure of the containment system does not exceed the design pressure duringtransportation [10 CFR 71.85(b)] as demonstrated by analysis (Appendix 2.10.1) and that there

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Y/LF-717/Rev I/page change 3/Ch-7/ES-3100 HEU. SAR/sc/06-30-07

are no pressure-relief devices [10 CFR 71.87(e)] in the package;

6. the external radiation levels for all transport conditions are within the allowable limits asmeasured for Normal Conditions of Transport (NCT) [10 CFR 71.87(j)];

,7. the nonfixed external contamination levels are within the allowable limits as demonstrated bysurface wipes prior to content insertion, containment vessel loading, and package closure[10 CFR 71.87(1)];

8. the contents are adequately sealed and have adequate space for expansion [10 CFR 71.87(d)];

9. all records for shipment areprepared and maintained; and

10. all lifting attachment features are either inoperative during transport [10 CFR 71.87(h)] or meetthe requirements of 10 CFR 71.45(a).

7.1.3.4 Leak Testing

Leak tests shall be conducted following the content loading and the containment vessel closure.The annulus between the 0-rings shall be leak tested toan acceptable leak rate of I X 10-4 ref-cm 3/s air(or equivalent) or lower in accordance with ANSI N 14.5-1997, Subclause 7.6.

7.1.3.5 Surveying

The radiation (gamma, neutron) emanating from the contents of the package shall be measuredbefore the package is released for transport [10 CFR 71.47 and 71.87(j)]. The package radiation doserate at the surface is measured to ensure that the content does not exceed the expected or allowable doserates. The package radiation dose rate at I m from the surface is measured to establish the TI for thepackage. The package exterior surface contamination level limits are found in 10 CFR 71.87(i) and49 CFR 173.443. The regulations present both fixed and nonfixed surface contamination level limits forthe various radionuclides. In addition to these limits, the user may have more stringent surfacecontamination levels that shall also be followed.

A final visual survey of the package and loading paperwork shall be conducted to ensure that thepackage was assembled correctly and that it is ready for final shipment preparation. This survey mayinclude a thorough review of the loading checklists by someone other than those who filled out the list toverify the loading Operations. The area immediately surrounding the assembly operations should besurveyed, and all spare or extra parts should be identified. A final package survey may include weighingthe package, hand-testing the closure nuts on the drum lid, and flexing the TIDs. The loading checklistshould include a place for this final quality check to be properly recorded-including a signature anddate-as being successfully completed.

7.1.3.6 Marking

The user shall ensure detailed marking procedures are consistent with 10 CFR 71185 and theapplicable subsections of 49 CFR 172, Subpart D. Each shipper shall ensure that each packagecontaining radioactive material is marked in the manner required.

Two electrochemically etched data plates are affixed to the exterior of the drum body in thelocations, and with the methods, indicated onDrawing M2E801580A031 (Appendix 1.4.8).

7-8


The packaging components (drum assembly, containment vessel body, lid, and closure nut) arealso marked with their serial numbers. The numbers are used to control these parts and to accumulatetheir respective histories.

7.1.3.7 Labeling

The user shall prepare labeling procedures that are consistent with the applicable subsections of49 CFR 172, Subpart E. The procedures should include the following steps, to ensure that:

1. the proper label is affixed to the package and the TI is determined at the time of loading;

2. the correct label (White-I, Yellow-Il, or Yellow-Ill) is determined using the table from49 CFR 1 72.403(c);

3. the appropriate label is affixed to two places on opposite sides of the drum;

4. the content name, nuclides and activity/mass (49 CFR 173.435), and the TI are entered in theblank spaces on the radioactive label; and

5. the information is entered legibly using a durable, weather-resistant means of marking.

Additionally, two Fissile labels are required per 49 CFR 172.402(d)(2). These labels must beaffixed two places on opposite sides of the drum adjacent to the radioactive labels. The CSI must belegibly entered on the Fissile label using a durable, weather-resistant means of marking.

7.1.3.8 Securing to Vehicle

The package shall be secured against movement within the vehicle in which it is beingtransported under conditions normally incident to transportation [49 CFR 177.834 and 177.842(d)]. Theloading procedures shall include the following measures, at a minimum, to ensure that:

1. only an approved conveyance is used,

2. all reasonable precautions are taken to prevent motion of the vehicle during loading,

3. no tampering with packages occurs during transit,

4. no vehicle is loaded or unloaded unless a qualified person is in attendance at all times, and

5. no radioactive material package is loaded onto a vehicle also carrying Div. 1.1 or 1.2 explosives(49 CFR 177.848).

7.2 PACKAGE UNLOADING

7.2.1 Receipt of Package from Carrier

Prior to shipment, the user shall verify that the receiver has agreed to accept the special nuclearmaterial. The user (shipper) shall ensure that appropriate documentation is submitted to the receiver toensure that the physical characteristics and hazards of the material are conveyed to the receiver.

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Y/LF-717/Rev I/page change 3/Ch-7/ES-3 100 H EU SAR/sc/06-30-07

The user shipping the package shall provide any special instructions to the receiver to safelyopen the package (10 CFR 71.89), including special tools and precautions for handling or unloading.These instructions shall include special actions in the event that TIDs are not intact, or if surfacecontamination or radiation surveys are too high.

The receiver shall accept the radioactive material by surveying the conveyance and packagesurface for contamination and external radiation levels. The receiver's procedures shall clearly indicatethat the contamination and radiation surveys and inspections be conducted upon receipt of the package.The receiver shall, at a minimum, include the following in their procedures (in compliance with10 CFR 71.111):

1. receive the package when offered by the carrier for delivery and,

2. monitor external surfaces of the conveyance and package for radioactive contamination andradiation levels.

All users shall include provisions for reporting safety concerns associated with the packaging orits use. The user shall notify NRC in accordance with and 10 CFR 20.2202. Incidents requiringnotification include:

1. removable radioactive surface contamination in excess of the limits provided by 10 CFR 71.87,and

2. external radiation levels in excess of the limits provided by 10 CFR 71.47.

The receiver shall compare the cargo with the list provided by the shipper. If a discrepancy appears 0between the cargo and the list, the receiver shall investigate and report to the NRC as required.

The package shall be removed from the conveyance prior to unloading the content. Unloadingprocedures shall, at a minimum, ensure that:

1. the package nonfixed surface contamination is below the minimum oil-site or off-siterequirements,

2. all appropriate package labels are affixed to the package exterior surface,

3. all lifting and handling equipment is certified for use,

4. all transfer equipment is certified for use,

5. the package is visually examined to ascertain surface damage that may have occurred duringshipping or handling, and

6. the TIDs are examined to ensure that the package has not been tampered with during shipment.

If the package surface was damaged during handling or shipping, a nonconformance tag shall becompleted and attached to the package for subsequent refurbishment (10 CFR 71.131). Ifthe TIDs arefound to be compromised, the receiver shall investigate and notify the NRC as required.

7-10


7.2.2 Removal of Contents

Detailed operating procedures shall describe activities required for content removal and shallidentify any safety and health measures required to protect workers and the environment. The proceduresshall include, at a minimum, the process steps listed below:

1. All appropriate labels for the material shipped are affixed to the exterior surface of the drumbody.

2. Surveys for nonfixed surface contamination and radiation dose rate measurements are conducted.

3. As applicable, nonfixed surface contamination in excess of limits is addressed as required.

4. The TliDs remain intact until removal.

5. The weld stud nuts and washers are removed and controlled.

6. The drum lid is removed.

7. Visible portions of the interior of the drum body and top plug are still in good condition-novisible signs of damage, water damage, or tears.

8. The top plug is removed using eye bolts that can be attached to the threaded holes provided onthe top plug.

9. The silicone rubber CV flange pad are removed from above the containment vessel.

10. The containment vessel top is in good condition-no visible signs of damage or loose closurenut.

11. A surface contamination check is conducted to discover any leak of radioactive material.

12. The containment vessel is removed from the drum assembly. The containment vessel is placedonto the work area. (This step may not be required if CV is unloaded while in the drum.)

13. The external retaining ring, containment vessel closure nut, and containment vessel sealing lidare removed and controlled. (No pressure buildup is expected under NCT.)

14. The 0-rings and the 0-ring grooves on the containment vessel flange are protected from damageduring unloading.

15. The HEU content (convenience cans or~bottle§) and associated packing materials (can spacers,stainless-steel scrubbers, silicone can pads, etc.) are removed from the containment vessel inaccordance with site-specific material-handling procedures.

16. The items removed and the inside of the containment vessel are checked for nonfixed surfacecontamination.

7.3 PREPARATION OF EMPTY PACKAGE FOR TRANSPORT

The user shall develop detailed procedures to prepare an empty package for storage or transport.

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Y/LF-717/Rev I/page Change 3/Ch-7/ES-3100 HEU SAR/sc/06-30-07

These procedures shall, at a minimum, ensure that:

1. The package has been emptied of all radioactive contents.

2. The radiation level at any point on the external surface of the package does not exceed0.5 mrem/h.

3. The nonfixed radioactive surface contamination on the external surface of the packaging does notexceed the limits specified in 49 CFR 173.443(a), and the internal contamination level does notexceed 100 times the limits in 49 CFR 173.443(a).

4. The package is not damaged, and there is no visible internal or external surface moisture orcorrosion.

5. The package is closed.

6. N6 leakage of radioactive material under conditions normally incident to transportation canoccur.

7. Any labels previously affixed in accordance with Subpart E of 49 CFR 172 are removed,obliterated, or covered. Leak-test labels should not be removed from the drum body.

8. The "EMPTY" label prescribed in 49 CFR 172.450 is affixed to the drum.

9. An appropriate notice is provided giving the name of the consignor or consignee. An examplenotice is, "This package conforms to the conditions and limitations specified in 49 CFR 173.428 0for radioactive material, excepted package-empty packaging, UN2908."

7.4 OTHER OPERATIONS

None.

07-12


SECTION 7 REFERENCES

10 CFR 71, Packaging and Transportation of Radioactive Material, Jan. 1, 2005.

49 CFR Pts. 100-180 and 393. Transportation, Oct. 1, 2004.

ANSI N 14.5-1997, Radioactive Materials-Leakage Tests on Packages for Shipment, American Nat!.Standards Institute, Feb. 5, 1998.

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Y/LF-717/Rev I/page change 3/Ch-7/ES-3 00 HEU SAR/sc/06-30-07


ATTACHMENT 3

MARK-UP OF CoC



NRC FORM 618 US. NUCLEAR REGULATORY COMMISSION

10cFR ! CERTIFICATE OF COMPLIANCEFOR RADIOACTIVE MATERIAL PACKAGES

a. CERTIFICATE NUMBER b REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES

9315 3 71-9315 USA/9315/B(U)F-96 1 OF 6

2, PREAMBLE

a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards setforth in Title 10. Code of Federal Regulations, Part 71, "Packaging and Transportation of Radioactive Material."

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation orpther applicable regulatory agencies, including the government of any country through or into which the package will be transported.

3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION

U.S. Department of Energy BWXT Y-12, L.L.C., application dated February 25,Washington, DC 20585 2005, as supplemented

4. CONDITIONS

This certificate is conditional upon fu]fi ling theequirvents of 10 CFR Part 71, as applicable, and the conditions specified below.

5.

.(a) Packaging

(1) Model No.: ES-3100

(2) Description , ....

The ES-3100 c ka t•• y ipproximately 110 cm (43 in) inoverall' heigh^.Nd 4 c R3 in o'Th - d is comv6ed of an outer drumassembly and ,r ipner conta mentTh tainmenfbessel is placed inside thedrum and surroundEd by a cement basbd.borated iieutron-absorber, Catalog 277-4. The

purpose of the Eg-_{00 is to transport bulk high enriched uranium in oxide form, uraniummetal and alloy, andl'uranyl nitrate crystals*

The outer drum assembly cohsistrof'•einf6rced stainless steel, standard mil spec 30-galdrum with an increased length. The volume formed between the drum and the attachedinner liner is filled with an inorganic, castable refractory material, Kaolite 1600TM, which iscomprised of concrete and vermiculite. The Kaolite 16 0 0 TM acts as both a thermal insulatingand an impact limiting material.

The containment vessel is approximately 82 cm (32 in) in overall height and 13 cm (5 in) inoverall diameter and is constructed of 304L stainless steel. The containment boundaryconsists of the 0.1 in thick containment vessel body and the lid assembly. The lid assemblyconsists of a sealing lid, a closure nut, and external retaining ring, which holds both theassembly and closure nut together. The double ethylene-propylene elastomer 0-rings in thetop flange of the containment vessel permit leak testing of the containment vessel. Themaximum gross weight of the package, including contents, is 190.5 kg (420 Ib).

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION18-20007 CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER C. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES

9315 3 71-9315 USA/9315/B(U)F-96 2 OF 6

5.(a) Packaging (continued)

(3) Drawings

The Model No. ES-3100 package is constructed and assembled in accordance with:

(i) BWXT Y-12, L.L.C., Drawing No. M2E801580A037, Sheets 1 through 6, Rev. A,"Consolidated Assembly Drawing."

(ii) BWXT Y-12, L.L.C., Drawing No. M2E801580A026, Rev. C, "Heavy Can SpacerAssembly."

(iii) Equipment Specification JS-YMN3-801580-AO01, Rev. E, "ES-3100 ContainmentVessel."

(iv) Equipment Specification JS-YMN3-801580-A002, Rev. "ES-3100 DrumAsser•Iy." 4k

(v) Equipment Speci-lIion JS-YMN3-801580.!O03, Rev. CVý.."Manufacturing ProcessSpecification for. catilgKaolite 1600. Ii( t the ES-3100 Shipping Package."

(vi) Equipment Spci ,JS-YMN3-8O'-A005 Rev. E, •'Casting Catalog No.277• N .bso qtb ES 4 Package.

5.(b) Contents (Type a orm I- tty of ierial perackage, andCriticality Safety Inde(CSI)I'- ,

The weight of the radi6a1 tije contents, co• ers, ca 'iR. attachments, polyethyleriebags, spacers, and othe r t•,al in the con1min bnt vessel shalirnt exceed 90 lb. The maximummass of hydrogenous packabi 'g materials in the containment'-vessel (e.g., polyethylene containersor bagging, silicone rubber pads, etc.) shall not exceed 500 grams. The maximum content decayheat load shall not exceed 0.4 watt§f. ..S.

The concentration limits of uranium and transuranic constituents shall, be the following:

Isotope Maximum Concentration

U-232 0.040 pg/gUa

U-233 0.006 g/gUb

U-234 0.02 g/gU

U-235 1.00 g/gU

U-236 . 0.40 g/gU

Transuranics (except Np) 40.0 pg/gU

Np-237 0.003 g/gUa pg/gU = 10- grams per gram of total uraniumb g/gU = grams per gram of total uranium

M A,-ý, W 4, P,, 5/I..

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION{8-;000•

o10 CFR 7 CERTIFICATE OF COMPLIANCESI FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER C. OCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES

9315 I 3 71-91 USA/9315/BUF 19 3 OF 6

5.(b) Contents (continued)

(1) Uranium as solid metal or alloy, packaged in stainless-steel or tin-plated carbon steelconvenience cans.

The. maximum uranium enrichment is 100 weight percent U-235.

For contents that must be shipped with spacers, the spacers must be in accordance withBWXT Y-12, L.L.C., Drawing No. M2E801580A026 and Equipment Specification JS-YMN3-801580-AO05, as specified in Condition No. 5,(a)(3). The quantity of fissile material in anyconvenience can shall not excded one-•hird of the mass loading limit per package for thatcontent. Spacers must be positioned between every two convenience cans.

(i) For metal and alloy in the form of, solid geometric shapes, meeting the followingrestrictio.Q mass limits are listed in Table 1. Conterntsn 'ot meeting the followingrestrictions must be shipped as broken metal (see Condition No. 5.(b)(1)(ii)).

(A) .. Spheresh.4• diameter no lareg6i4r 3.2 n(maximu osphers per conyv.ince can) "

(B) Cylinders It a diamneter.no larithan 3.24 in,,• (maximum-, oyhnder per !y,, irence can)

(C) Squ`•• .,ha'iri ion no 16 rOr than 2.29 in x 2.29 in

(D) ha Ing rnen6io f2'55 5 meter x 2ln0allIrpmum of 1 slizs 1cove c an)-,

Table 1: Loading, Lifrits for Metal and Alloy in Solid. Geometric Shapes

Solid uranium Uranium -WitihSpacers No Spacersmetal or alloy Enrichment CSI Mai)imum Mass U-235 Maximum Mass

(specified (weight (kg) U-235 Pergeometric shapes) percent U-235) Per Convenience Per Package Package (kg)

Can

Spheres _ 100 0.0 0.000 0.0 0.0

Cylinders • 100 0.0 6.000 18.000 12.000

Sq. Bars !5 100 0.0 1 10.000• 30.000 18.000

Slugs > 80 0.0 5.447 16.342 Spacer req'd

Slugs • 80 0.0 8.738 26.213 Spacer req'd

I mtk'to 4 Z'-.r.1618 U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCEFOR RADIOACTIVE MATERIAL PACKAGES

A. CERTIFICATE NUMBER

9315b. REVISION NUMBER

3C. DOCKET NUMBER

71-9315d. PACKAGE IDENTIFICATION NUMBER j PAGE

USA/9315/B(U)F-96 4

PAGES

6. OF'1

5.(b)(1) Contents (continued)" ta \ T,""'E-IT ,

(ii) For metal or alloy defined as broken metal, mass limits are specified in Table 2. Uranium Jmetal and alloy pieces must have a surface-area-to-mass ratio of not greater than 4,1.00 cm2/g or must have a mass not less than 50 g, whichever is most restrictive. Powders,foils, turnings, wires, and incidental small particles are not permitted, unless they arerestricted to not more than 1 percent by weight of the content per convenience can, and theyare either in a sealed, inerted container or are stabilized to an oxide prior to shipment.

Table 2: Loading Limits for Solid Metal or Alloy in the Form Defined as Broken Metal

Uranium Enrichment With Spacers No Spacers(weight percent CSI Maximum Mass Maximum Mass

U-235) U-235 (kg)8 U-235 Per Package.... (kg)a3

Per Convenience (g'_ ... i - Can P te Package .... _ __:_ _ _ _

> 95 and :5 100 -;.....0.0 2>'nd"0000-• 0.,925 . 2.774 Spacer req'd

0.4 .4 .. 849, .-. 5548 Spacer req'd

0.8.,. 873.23, Spacer req'd

.0,, '7 . Spacer req'd> 90 and s 95 " 637 • Spacer req'd

.. 5.274 -Spacer req'd

0.8" 3.516 1-.., Spacer req'd

2.0 .. 568 16.703 Spacer req'd

> 80 and 90 2.500 Spacer req'd

0.4 2.500 7.500 Spacer req'd

0.8 3.333 10.000 . Spacer req'd

2.0 5.278 15.834 Spacer req'd

> 70 and • 80 0.0 0.742 2.225 2.225

0.4 2.967 8.900 4.450

0.8 0.000 0.0 0.0

2.0 7.911 23.734 0.0

USA/93151B(U)F-96 Mark up for Rev 5

INSERT B

(ii) For metal and alloy defined as broken metal, mass limits are specified in Table 2.Uranium metal and alloy pieces must have a surface-area-to-mass ratio of not greaterthan I cm 2/g or must not pass freely through a 3/8-inch (0.0095m) mesh sieve. Particlesand small shapes that do not pass this size restriction, as well as powders, foils, turnings,and wires are not permitted, unless they are either in a sealed container under an inertcover gas or are completely stabilized to an oxide prior to shipment.. Uranium metal oralloy which has been stored in water must be shipped in a sealed container under an inertcover gas or completely oxided prior to shipment.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION10 CFR 71 CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGESI1. a. CERTIFICATE NUMBER b. REVISION NUMBER . DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES

9315 3 7 315 USA/9315/B(U)F-96 5 OF 6

5.(b)( 1) Contents (continued)

Table 2: Loading Limits for Solid Metal or Alloy in the Form Defined as Broken Metal (Continued'

(2) Uranium as o~e, wl' 30 pa,,,ged in stainless-steel, tin-plated carbon "eel, or icke i1oy ..ih)ce ,u or polygU~ylene bottles. The physicalform of all cont~ptjs dense loosa 'wde)-.Whic y corn t11nlumps and pellets. Moisturecontent in oxide is'jpi t d to 3 weight' jerent water. The 4ss limit shall be 24.0 kg of

oxide, with a maximu _. nass of 21.124 kg U-235, with"t,)ESI of 0.0. No spacers arerequired in the containmentvyessel.

(3) Solid uranyl nitrate in the form of uranyl nitrate crystals, [U0 2(NO3)2.xH 20, where x is • 6].Uranyl nitrate crystals must be contained in a non-metallic convenience container (such aspolyethylene bottles). The mass limit shall be 0.0 kg of uranyl nitrate crystals, with a

-maximum mass of 0.0 kg U-235, with a CSi of 0.0. No spacers are required in thecontainment vessel.

6. The vent holes on the outer steel drum shall be capped closed during transport and storage topreclude entry of rain water into the insulation cavity of the drum..

7. Content forms may not be mixed in a single ES-3100 containment vessel.

NRC FORM'618 U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCEFOR RADIOACTIVE MATERIAL PACKAGES

a CERTIFICATE NUMBER b REVISION NUMBER c. DOCKET NUMBER d PACKAGE IDENTIFICATION NUMBER PAGE PAGES

9315 I 3 71-9315 USA/9315/B(U)F-96 6 OF

8. Any combination of convenience can sizes is allowed in a single package, as long as, the total heightof the can stack (including silicone rubber pads and spacers, if required) does not exceed the insideworking height of the containment vessel (31 in). Any closure on the convenience can is allowed.

9. Empty convenience cans, spacers, silicone rubber pads, and/or stainless-steel scrubbers (i.e.,stainless steel trimmings that act as dunnage) may be used to fill the void space in the containmentvessel. Empty convenience cans must have a minimum 0.125 in diameter hole through the lid.

10. The contents and the convenience cans may be bagged or wrapped in polyethylene for

contamination control provided the limits of Condition No. 5.(b) are met.

11. Transport by air is not authorized.

12. In addition to the requirements of Subpart G of 10 CFR Part 71:

(a) The package shall be prepared for shipment and operated in accordance with the PackageOperations in Section 7 of the application, as supplemented.

(b) Each package must meet the Acceptance Tests and Maintenance Program of Section 8 of theapplication, as supplemented.

13. The package authorized by this certificate is hereby approved for use under the general licenseprovisions of 10 CFR 71.17.

14. Expiration date: April 30, 2011.

REFERENCES

BWXT Y-12, L. L.C., application dated February 25, 2005, as supplemented.

BWXT Y-12, L. L. C., supplements dated April 27, May 26, August 15, 2005; and January 9, February 6,March 20, May 8, June 6, July 18, August 21 and 24, October 26, 2006; and January 19 and April 26, 2007.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION

Robert A. Nelson, ChiefLicensing BranchDivision of Spent Fuel Storage and TransportationOffice of Nuclear Material Safety

and Safeguards

Date: May 8, 2007

Washington, JUN 2 7 2007 - Nuclear Regulatory Commission

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Transcript of Washington, JUN 2 7 2007 - Nuclear Regulatory Commission