GE Nuclear Energy Vallecitos Nuclear Center Sunol, CA ... · [1.2] Model 2000 Radioactive Material Transport Package 2000 Watts Decay Heat Upgrade Safety Analysis Report, NEDO-32318,

GE Nuclear Energy

Vallecitos Nuclear Center6705 Vallecitos RoadSunol, CA 94586925-862-4344

March 25, 2005CM05007

John D. Monninger, ChiefLicensing SectionSpent Fuel Project OfficeOffice of Nuclear Material and SafeguardsWashington, D.C. 20555-0001

Attention: Document Control Desk

Reference: 1) USNRC Certificate of Compliance No. 9228, Docket Number 71-9228.2) NEDO-31581, "Model 2000 Radioactive Material Transport Package Safety Analysis

.Report', dated April 1988.3) NEDO-34408, "Model 2000 Radioactive Material Transport Package MTR-Type Divider

and Tower Shielding Reactor Fuel Basket Safety Analysis Report', dated July 1994.

Dear Mr. Monninger:

Enclosed are six copies of our application, pursuant to 10CFR71.19(e), for revision of theModel 2000 shipping container identification number to include the suffix "-96" and to anamendment to the Certificate of Compliance (1) to allow the transportation of the package inthe horizontal mode for LWR spent fuel and hardware/radioisotope material contents. Theapplication is presented in the form of replacements pages to the Safety Analysis Reports (2)and (3) included in Attachments A through C. Changes within these pages are indicated by avertical line in the outer edge of the page.

To support the "-96" designation, we included herein a table addressing the nineteen issuesconsidered in the rulemaking process that resulted in the revised rule and what changes, ifany, has each issue on the Model 2000 package. In addition, we provide a new Appendix forReference (2), Attachment C, describing the shielding analyses performed to demonstratecompliance to the radiation level limits of the regulation when the package is transported in thehorizontal mode.

CM05007Page 2

Issue Resolution Changes

1. Changing Part 71 to the International Proposal not adopted In final rule. No changes are needed.Systems of Units (SI) Only

2. Radionuclide Exemption Values Adopted radionuclide activity Issue Is not applicable based on theconcentration values and consignments Model 2000 design and authorizedactivity limits in TS-R-1 and exemption contentfor certain natural materials and ore

3. Revision of Al and A2 Adopted changes In the Al and A2 Issue Is not applicable based on thatvalues from TS-R-1, with the exception the Model 2000 is a leak tight containerof two radionuclides. as demonstrated in Chapter 4 of the

SAR.

4. Uranium Hexafluoride (UFO) Adopted. Issue Is not applicable, since thePackage Requirements. package Is not authorized for the

transportation of UFO.

5. Criticality Safety Index (CSI). Adopted the CSI requirements from TS- Revised Chapters 1 and 6 of the SAR'sR-1. to Incorporate CSI nomenclature. See

Attachment A

6. Type C Packages and Low Proposal not adopted In final rule. No changes are needed.Dispersible Material.

7. Deep Immersion Test Adopted as an extension of the previous Previously submitted Informationversion of 10 CFR 71.61 (February 1998) demonstrated

compliance to this rule. Table 1.3.1.1has been changed to clarify compliance

._ with this rule. See Attachment B.

8. Grandfathering Previously Approved Adopted a process for allowing This submittal provides InformationPackages continued use, for specific period of demonstrating compliance with the final

time. rule, no Grandfathering Is requested

9. Changes to Various Definitions. Adopted several revised and new No changes are needed.definitions.

10. Crush Test for Fissile Material The revised 10 CFR 71.73 expanded The mass of the Model 2000 exceedsPackages. the applicability of a crush test to fissile 500 kilograms (1100 pounds), therefore

material packages with a mass not this rule Is not applicable.greater the 500 kilograms (1100pounds).

11. Fissile Material Package Design for Adopted packaging design The Model 2000 Is not transported byTransportation by Aircraft requirements for packages transporting air, for fissile material contents.

fissile material by air.

12. Special Package Authorization. Adopted provisions for special package Not applicable to Model 2000.authorization to transport largecomponent, under specialcircumstances and only to one-timeshipment.

13. Expansion of Part 71 Quality Expanded the scope of Part 71 QA ChapterI of the Model 2000 SAR'sAssurance (QA) Requirements to requirements to apply to any person have been changed to explicitly indicateCertificate Holders. holding or applying for a Certificate of the GE QA program compliance to 10

Compliance. CFR 71.101 (a), (b) and (c). SeeAttachment B.

14. Adoption of the American Society Not adopted. No changes are needed.of Mechanical Engineers (ASME) code.

15. Change Authority for Dual-Purpose Not adopted. No changes are needed.Packages Certificate Holders.

16. Fissile Material Exemptions and Adopted various revisions to the fissile Chapter 6 of the SAR's demonstrated

CM05007Page 3

General Lcense Provisions. material exemptions and the general that criticality safety of the Model 2000license provisions In Part 71. does not rely on limiting the fissile

materials to exempt or generallylicensed quantities. No changes areneeded.

17. Double Containment of Plutonium. Adopted rule removed the double The Model 2000 SAR's has nocontainment requirement of plutonium reference to the double containment ofmaterial in quantities greater than 0.74 Pu materials. No changes are needed.terabecquerel (20 curies).

18. Contamination Limits as Applied to Not adopted No changes are needed.Spent Fuel and High Level Waste.19. Modification of Events Reporting Adopted rule modified reporting While the final rule Is applicable to theRequirements. requirements. package, no changes are needed to the

SAR's to conform to the new rule..

Please contact the undersigned if you require additional information.

Very truly yours,

David W. TurnerE-mail: david.turner(cgiene.ge.com

Attachments

cc: C. Martinez, GEL. Quintana, GER. Pomares, GE

CM05007Page 4

ATTACHMENT A

Replacement Pages for Reference (3)

NEDO-32408 March 2005REVISION 2

e) "MTR-type fuel elements including the TSR fuel with a maximum decay heat of 1500.watts. Fissile content not to exceed 710 grams of U235 per divider position."

In this revision of the Safety Analysis Report, authorization is sought for the use of the MTR-type fuel divider to transport several types of TRIGA fuels and MTR-type fuels with enrichmentup to 94% and the addition of MTR-type Umetaj fuel meat composition with 95% enrichment.Previously, authorization was sought (Revision la) for TRIGA fuel and MTR-type fuels withenrichments up to 93.2%, 120 days minimum decay time, and a Criticality Safety Index (CSI) ofzero (0.0).

The MTR-type generic fuel element used in the evaluation of the divider is primarily analuminum construction assembly of composite fuel plates. The individual fuel plates arecomposed of a fuel matrix clad with aluminum. The fuel elements contain up to a total of 355gm U235 with an enrichment up to 94%. The maximum U235 content per shipment, based oncriticality limit, is 14,910 grams for U30 8 fuel, 14,595 grams for U3 Si2 fuel, 23,352 grams forUAI, fuel, and 4620 grams for Umetal fuel for a CSI of zero (refer to Chapter 6 for details). Themaximum weight of each assembly, including the fuel and all required hardware, is not to exceed42.8 lb. per divider cell. The minimum cooling time of each assembly is 120 days with themaximum decay heat generation per fuel cell of 120 watts maximum with a distribution of 85watts maximum for the lower half of the fuel cell and a maximum of 35 watts in the upper half ofthe fuel cell. The maximum decay heat per shipment is limited to 1500 watts.

Table 1.1 provides dimensions and nuclear data for several MTR-type fuel which are envelopedby a generic element used for this analysis.

The TRIGA fuel assemblies consist of cylindrical zirconium hydride fuel elements with orwithout graphite reflectors within an aluminum, stainless steel, or inconel cladding. Figures 1.7and 1.8 show a typical TRIGA fuel elements. Table 1.2 provides dimensions and nuclear datafor some typical TRIGA fuel elements. The TRIGA fuel elements are transported within thesame divider as the MTR type fuel-assemblies. Multiple TRIGA fuel elements may betransported in each divider cell. The configuration is limited by geometry and U235 content of theTRIGA fuel elements. The maximum U235 content of the TRIGA fuel, based on the criticalitylimit, in each of the 21 divider cells shall not exceed 1370 grams for enrichment up to 70%(28,777 grams per shipment), 1219 grams per cell for up to 94% enrichment with graphitereflector (25,599 grams per shipment), and 494 grams per cell for up to 94% enrichment withoutgraphite reflector (10,374 grams per shipment), all with a. CSI of zero (refer to Chapter 6 fordetails). The minimum cooling time of each assembly is 120 days, with the same decay heatlimits per fuel cell and per shipment as for the MTR fuel. The maximum allowable weight of theTRIGA fuel assemblies including shoring devices shall not exceed 42.8 lb. per divider cell. Themaximum weight of the MTR-type fuel divider, loaded with TRIGA fuel assemblies, shall notexceed 5450 lb., the maximum allowable payload of the Model 2000 package.

The TSR fuel core consists of 25 sub-assemblies of aluminum-clad fuel matrix assembled toform a sphere. The total fuel loading of the TSR fuel core assembly is 8451 gm of U235. TheTSR fuel core, the only one in existence, with four additional sub-assemblies (lune platescurrently in storage), shall be transported in a minimum of two shipments.

1-2


1.3 REFERENCES

[1.1] Model 2000 Radioactive Material Transport Package Safety Analysis Report,NEDO-31581, April, 1988.

[1.2] Model 2000 Radioactive Material Transport Package 2000 Watts Decay Heat UpgradeSafety Analysis Report, NEDO-32318, July, 1994.

[1.3] Code of Federal Regulations, Title 10, Part 71, 2004.

[1.4] General Electric Quality Assurance Program for Shipping Packages for RadioactiveMaterial, QAP-1, Revision 5, 1990.

1.4 APPENDIX

1.4.1 Drawings

105E9557 Fuel Divider Certification Drawing

105E9560 TSR Fuel Basket Certification Drawing

1-21


3. Accident Array - Twice the allowable number of packages (2 x N) must remain subcriticalif the packages are subjected to hypothetical accident conditions and are stacked together inany arrangement, closely reflected on all sides by water, and with optimum interspersedhydrogenous moderation. In the accident condition, no significant reduction in volume(deformation) of the cask is expected. Also, the contents of the cask are expected to remaindry. For N-i, the accident array demonstration is a subset of the normal condition array.

A conservative demonstration of subcriticality for these three scenarios qualifies the Model 2000cask (with TRIGA and MTR type fuel) as a Fissile Class III package for transport within theUnited States. The above criteria are consistent with those stated in Sections 71.55 and 71.61 of1 OCFR Part 71.

6.1.1.2 Infinite Cask Scenario

In order to achieve a zero Criticality Safety Index (i.e., an infinite number of cask shipments), aslight modification to the previously approved GEMER methodology is being proposed. Thismodification is described and validated in Appendix 6.7.4. The application of the Infinite CaskMethodology qualifies the cask loading as an infinite (N=o) Fissile Class III package with a zeroCriticality Safety Index under the provisions of 1 OCFR71.61. This loading also meets the IAEASafety Standards for an infinite shipment (N=-) with a zero Criticality Safety Index.

6.1.2 Container Description

The Model 2000 cask is shown in Figure 6.1. The MTR type of fuel divider structure is shown inFigure 6.2. The support structure is essentially a solid cylindrical "core" of 304 stainless steelwith 3.15 by 3.50 inch slots (8.00 by 8.89 cm) for fuel elements. Single or multiple MTR typefuel elements (stacked end-to-end) fit in each slot in the grid. The divider structure is also usedto ship TRIGA fuels. The support assembly fits inside the 26.5 inch (67.31 cm) diameter cavityof the Model 2000 cask.

The Model 2000 MTR type fuel divider design incorporates neutron poison material to ensurecriticality safety. Borated aluminum sheets are built into the fuel support structure, providing a"neutron curtain" between fuel elements. The Boron content in the divider is conservativelymodeled by reducing the boron atom density to 60% of the minimum specification.

6-2

- -

NEDO-32408REVISION 2

March 2005 |

(111"!-1,71Table 6.13. Summary of Results - Infinite Cask Scenario for TRIGA Case VIIIType of Fuel (94% Enriched)

Condition/Parameter Quantity/Comment

Number of Containers in Demonstration Infinite Array

Optimum Hydrogenous Moderation (In Fuel) 0.200 Weight Fraction H20

Maximum Multiplication Factor 0.92156kff + 2a + Bias

6.2 PACKAGE FUEL LOADING

6.2.1 MTR

Table 6.14 and Table 6.15 described herein qualifies as a single (N=1) Fissile Class III packageunder the provisions of 1OCFR71.61. This loading also meets the IAEA Safety Standards for asingle shipment (N=1) with a Criticality Safety Index of 50..

Tables 6.16 through 6.19 summarize the maximum fuel load (mass), as well as some significantconstraints for MTR fuel cases I, II, III, and IV, respectively, as a result of the infinite caskanalysis.

As the Infinite Cask Methodology is applied to the GEMER analyses, the loading specified inTables 6.16 through 6.19, described herein, qualify as an infinite (Noo) Fissile Class III packagewith a zero Criticality Safety Index under the provisions of 1OCFR71.61. This loading alsomeets the IAEA Safety Standards for an infinite shipment (N=co) with a zero Criticality SafetyIndex. The loadings specified in Tables 6.16 through 6.19 envelope those specified in Tables6.14 and 6.15.

I

Table 6.14. Model 2000 235U Mass Limit for a Single (N=1) Shipment for Case I (U308 )Type of Fuel (93.2% Enriched)

Number of Allowable Basket 235U Mass Per Position Total 235U MassPositions (gm) (gm)

21 710 14,910

6-15

.


6.2.2 TRIGA

Table 6.26 provides dimensions and nuclear data for several TRIGA type fuels to which thisanalysis applies.

6.2.2.1 TRIGA Case XVII

The GEMER model for the TRIGA Case XVII fuel consists of 10 fuel elements loaded into eachMTR divider position. A total of 1370.0 grams of 70% enriched U-235 per divider position inthe MTR type fuel divider is modeled in this criticality analysis. This represents 10 TRIGA fuelelements per divider position. The corresponding masses of U-238, and cladding material can beestimated. The detailed calculations of these number densities are shown in Subsection 6.7.2.1.2.

Table 6.25 summarizes the maximum fuel load (mass), as well as some significant constraints forthe TRIGA Case XVII type fuel. As the Infinite Cask Methodology is applied to the GEMERanalyses, the loading specified in Table 6.25 described herein qualifies as an infinite (N=o)Fissile Class III package with a zero Criticality Safety Index under the provisions of IOCFR71.61. This loading also meets the IAEA Safety Standards for an infinite (N=co) shipment with azero Criticality Safety Index.

Table 6.26 provides the packaging parameters for the Model 2000 cask with TRIGA Case XVIItype fuel.

Table 6.25. Model 2000 235U Mass Limit for an Infinite (Ncio) Shipment for TRIGA Case XVIIType of Fuel (70% Enriched)

Number of Allowable Basket 235U Mass Per Position Total 235U MassPositions (gm) (gin)

21 1370.0 28,770

6-22


March 2005 l

Table 6.27. Package Parameters for the Model 2000 Cask with TRIGA Case XVIIType of Fuel (70% Enriched)

Parameter Quantity/Comment

Fuel Construction Cylinders

Fuel Material UZrH1.6 (8.5 wt% U)

Cladding Material Stainless Steel

Maximum Fuel Enrichment 70%

Minimum Fuel Bumup N/A

Burnable Poison Materials 20 to 36 grams per fuel element

Nominal Stainless Steel Density 7.827 g/cc

6.2.2.2 TRIGA Case V

The GEMER model for the TRIGA fuel consists of 1219.0 grams of 94% enriched U-235per divider position in the MTR type fuel divider. This represents 7 TRIGA fuelelements loaded into each divider position. Based on the U-235 mass (174.143 grams perfuel element) provided in Table A.1-2 of Reference [6.2], the corresponding masses ofU-238, and cladding material can be estimated. The detailed calculations of these numberdensities are shown in Subsection 6.7.2.2.2.

Table 6.28 summarizes the maximum fuel load (mass), as well as some significantconstraints for the TRIGA Case V type fuel. As the Infinite Cask Methodology is appliedto the GEMER analyses, the loading specified in Table 6.28 described herein qualifies asan infinite (N=oo) Fissile Class III package with a zero Criticality Safety Index under theprovisions of lOCFR 71.61. This loading also meets the IAEA Safety Standards for aninfinite (N=oo) shipment with a zero Criticality Safety Index.

Table 6.29 provides the packaging parameters for the Model 2000 cask with TRIGA CaseV type fuel.

Table 6.28. Model 2000 235U Mass Limit for an Infinite (N=co) Shipment forTRIGA Case V Type of Fuel (94% Enriched)

Number of Allowable Basket 235U Mass Per Position Total 23 'U MassPositions (gm) (gm)

21 1219.0 25,600

6-24


March 2005 |

Table 6.29. Package Parameters for the Model 2000 Cask with TRIGA Case VType of Fuel (94% Enriched)



Fuel Material UZrH1. 6 (8.5 wt% U)


Maximum Fuel Enrichment 94%

Minimum Fuel Burnup N/A

Burnable Poison Materials None


6.2.2.3 TRIGA Case VIII

The GEMER model for the TRIGA fuel consists of 493.9 grams of 94% enriched U-235per divider position in the MTR type fuel divider. This represents 12 fuel elements perdivider position. Based on the 235U mass (41.16 grams per fuel element) provided inTable A.1-2 of Reference [6.21], the corresponding masses of 238U and cladding materialcan be estimated. The detailed calculations of these number densities are shown inSubsection 6.7.2.3.2.

Table 6.30 summarizes the maximum fuel load (mass), as well as some significantconstraints for the TRIGA Case VIII type fuel. As the Infinite Cask Methodology isapplied to the GEMER analyses, the loading specified in Table 6.30 described hereinqualifies as an infinite (Nco) Fissile Class III package with a zero Criticality SafetyIndex under the provisions of 1OCFR 71.61. This loading also meets the IAEA SafetyStandards for an infinite (N=oo) shipment with a zero Crity Cality Safety Index. Table6.31 provides the packaging parameters for the Model 2000 cask with TRIGA Case VIIItype fuel.

Table 6.30. Model 2000 235U Mass Limit for an Infinite (Nco) Shipment for TRIGACase VIII Type of Fuel (94% Enriched)

Number of Allowable Basket 23sU Mass Per Position Total 235U Mass

Positions (gm) (gm)

21 493.9 10,370

6-25

-

NEDO-32408 March 2005 |REVISION 2

[6.12] Tuck, G. and I. Oh, "Benchmark Critical Experiments on Low-Enriched UraniumOxide Systems with H/U=0.77", Rockwell International: NUREG/CR-0674,August, 1979.

[6.13] Bierman, S.R., M. Durst, and E.D. Clayton, "Criticality Experiments withSubcritical Clusters of 2.35 wt% 235-U Enriched U0 2 Rods in Water withUranium or Lead Reflecting Walls", Batelle Pacific Northwest Laboratory:NUREG/CR-0796, April, 1979.

[6.14] Dabbs, R.D., "HFIR Spent Fuel Rack Criticality Safety Analysis", ORNL/C-HFIR-91-021, 4 October, 1991.

[6.15] Hardy, J., Jr., et al., "Nuclear Science Engineering", 40, 101. 1970.

[6.16] Bohn (ed.), E.M., et al., "Benchmark Testing on ENDF/B-IV", ENDF-230, Vol. 1.March, 1976.

[6.16] "1 OCFR (Code of Federal Regulations - Energy), Part 71 ", Revised as of1 October, 2004.

[6.17] "Regulations for the Safe Transport of Radioactive Material", IAEA. 1996 Ed.,Vienna, 2003.

[6.18] Gwin, J.C., et. al., "The Measurement of Eta and Other Nuclear Properties ofU233 and U235 in Critical Aqueous Solutions", Nuclear Science Engineering, 12.364. 1962.

[6.19] Wagner, J.C., et. al., "MCNP: Criticality Safety Benchmark Problems",Los Alamos National Laboratory: LA-12415, October, 1992.

[6.20] Application for Revalidation of Certificates of Compliance and CompetentAuthority Certifications (USA/9034/AF and USA/9037/AF) for TRIGAI andTRIGA2 Shipping Packages, General Atomics, June 10, 1992.

[6.21] Ravnik, M., and Glumac, B., "TRIGA Spent-Fuel Storage Criticality Analysis",Nucl. Tech., 114, 365, 1996.

[6.22] Mele I., et. al., "TRIGA MARK II Benchmark Experiment, Part I: Steady-StateOperation", Nucl. Tech., 105, 37, 1994.

[6.23] Mele I., et. al., "TRIGA MARK II Benchmark Experiment, Part II: PulseOperation", Nucl. Tech., 105, 52, 1994.

6-68


March 2005 l

Table 6.57. Summary of Results - Accident Condition (Conservatively Treatedas Flooded) for the TSR Type of Fuel (93.2% Enrichment)


Number of Containers in Demonstration 2 (Side-By-Side)

Optimum Hydrogenous Moderation (In Fuel) 0.50 Weight Fraction H2O

Optimum Hydrogenous Moderation 0.1 Weight Fraction H20 (Between Casks)(Interspersed Moderation) 0.1 Weight Fraction H20 (Inside Casks)

Reflector 30.5 cm (I ft) Full Density H20

Maximum Multiplication Factor 0.91500kOff + 2a + Bias

Table 6.58. Summary of Results - Infinite Cask Scenario for TSR Type Fuel(94% Enruchment)



Optimum Hydrogenous Moderation (In Fuel) 0.50 Weight Fraction H20

Maximum Multiplication Factor 0.92694krff+ 2a + Bias

6.7.3.2 Package Fuel Loading

Table 6.59 summarizes the maximum fuel load (mass), as well as some significantconstraints for the form of the TSR type of fuel.

The loading specified in Table 6.60 described herein qualifies as an infinite (N=oo) FissileClass III package with a zero Criticality Safety Index under the provisions of 1 OCFR71.61based on the infinite cask analysis. This loading also meets the IAEA Safety Standardsfor an infinite shipment (N=oo) with a zero Criticality Safety Index. Table 6.61 shows thepackage parameters for the TSR fuel shipment.

I

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WATER MASS (GRAMS) = 120000.0TOTAL VOLUME(FUEL+CLAD+WATER)= 134337.1 CMA3

CONCENTRIC CYLINDER CASECYLINDRICAL HEIGHT = 38.46936 CMHZ( 3) = -2.274357 CMINNER CONCENTRIC RADIUS = O.0000000E+00 CMOUTER CONCENTRIC RADIUS = 33.34000 CMN1 N16 N131 N238 N235

1 5.9770033E-0216 2.9885016E-02

131 6.3964571E-032381 1.6972011E-062351 2.6929632E-05

TOTAL 9.6080132E-02H/U-235 RATIO = 2219.489MIXTURE DENSITY = 1.191033 G/CCOVERALL H/U-235 RATIO = 1558.965

6.7.4 Validation of Infinite Cask Methodology (1CM)

In order to achieve a zero Criticality Safet Index (i.e., an infinite number of casks) from acriticality stand point, we have applied the NRC approved GEMER methodology toanalyze an infinite array of Model 2000 shipping casks.

A unit cell consisting of the cask surrounded by water in a square pitch is chosen forthe analysis. The sides of the cell are fully reflecting, thus making this system aninfinite array of identical unit cells. One of the definitions of the multiplication factor ofa system is:

k = net neutron production ratenet neutron loss rate due to capture and leakage (1)

In an infinite array, the neutron leakage is zero; therefore, all the loss from the system isas a result of absorption reactions within the system. The kef in this case is denoted bythe infinite multiplication factor ko. Thus if k. is less than 1, the system will nevergo critical.

6.7.4.1 Methodology Validation

The MTR Case I fuel (U30s) with 93.2% enrichment is used as an exampledemonstration for this ICM. From the previous submittal, the k-effective from the threescenarios (normal, single, and accident arrays) are shown in Tables 6.73 through 6.75.The maximum keff + 2o + Bias in this example is 0.93559 for the accident array as shownin Table 6.75. All three scenarios (normal, single, and accident arrays) have one-footthick rectangular water reflectors on all sides. As a result of the limited U-235 mass(710 grams per divider slot and 14910 grams total in each cask), as well as the presenceof boral plates between the divider slots, the cask is designed to be subcritical under allnormal and accident conditions, as shown by the keff + 2oa + Bias in Tables 6.73 through

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6.7.4.2 Discussion and Conclusions

From Figure 6.45, as the edge-to-edge distance increases for a fixed water density, thek-effective decreases as a result of the greater neutron absorption in the interstitial water.For the totally-leaking configuration, the k-effective is the smallest. This is not surprisingas any neutrons leaking out of the system will not return. The totally leaking system isabout 3% (0.91462) lower in reactivity than the leak-proof case (0.94519) at zero waterdensity.

The maximum keff + 2o + Bias from Tables 6.73 through 6.75 is 0.93559. This numbershould be less than the leak-proof case (0.94519), which validates the ICM.

Also, as the water density increases for a fixed edge-to-edge distance, the k-effectivedecreases as a result of neutron absorption in the interstitial water.

This section clearly shows that by applying the ICM to any cask loaded with any fuelelements, we have covered an infinite array of identical shipping casks loaded with thesame fuel elements. This justifies the use of a zero Criticality Safety Index in Model 2000shipping cask shipments if the ICM is applied in the GEMER calculations.

6.7.5 Criticality Evaluation (Discrete Model) for MTR-Fuel Divider SN 01(DELETED)

6.7.6 Criticality Evaluation (Discrete Model) for MTR-Fuel Divider SN 01

6.7.6.1. Discussion and Results

6.7.6.1.1 General

The Model 2000 cask, as discussed earlier in Subsection 6.1.2, is a cylindrical cask withlead shielding. The cask has a cavity that is 26.5 inches inside diameter and 54 incheshigh for holding nuclear radioactive material, including special nuclear material.

This application serves two purposes:

First, this application applies the discrete fuel modeling approach to the MTR Type 1, II,III, and IV fuels. This represents a more realistic approach in modeling the MTR Type I,II, III, and IV fues and is different from the homogenous fuel model discussed earlier inSection 6.4.3. The NBSR fuel was found to be the most limiting U308 -based fuels andwas chosen to envelope the MTR Type I fuel study. Similarly, the ORR fuel was chosento envelope the U3 Si2 -based MTR Type II fuel and the SAPHIR fuel was chosen toenvelope the UAl-based MTR Type III fuel. MTR Type IV fuel was chosen to envelopethe metallic form of the MTR fuel.

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NEDO-32408 March 2005 lREVISION 2

Table 6.100. Summary of Results for the Infinite Cask Scenario for MTR Type IV Fuel(95% Enrichment) -Discrete Fuel Model



Optimum Hydrogenous Moderation 1.0 Weight Fraction H20

Maximum Multiplication Factor 0.62514kOff+ 2a + Bias


Tables 6.101 through 6.104 summarize the maximum fuel load (mass), as well as somesignificant constraints for the form of the MTR Type I, II, III, and IV fuel, respectively.The loading specified in Tables 6.101 through 6.104 described herein qualifies as aninfinite (N=oo) Fissile Class III package with a zero Criticality Safety Index under theprovisions of 1OCFR71.61, based on the infinite cask analysis. This loading also meetsthe IAEA Safety Standards for an infinite shipment (N=oo) with a zero Transport Indexfor criticality safety.Criticality Safety Index.

Table 6.101. Package Parameters for Model 2000 Cask with MTR Type I Fuel (U3 08 )


Fuel Construction MTR Type Fuel Plates

Fuel Material U 3 0 8 (41.6%) + Al (58.4%)

Cladding Material Al

Minimum Fuel Bumup N/A

Total 235U mass per divider position 710 grams

Maximum Fuel Enrichment 94.0%

238U mass per divider position 45.3 grams

Burnable Poison Materials (e.g., Gd) None

Maximum U308 Density (100% T.D.) 8.39 gm/cm3

Nominal Al Density 2.699 gm/cm 3

6-285



Tables 6.122 through 6.124 summarize the maximum fuel load (mass), as well as somesignificant constraints for the form of the TRIGA Case XVII, V, and VIII fuel,respectively. The loading specified in Tables 6.122 through 6.124 described hereinqualifies as an infinite (N=oo) Fissile Class III package with a zero Criticality Safety Indexunder the provisions of I OCFR71.61 based on the infinite cask analysis. This loadingalso meets the IAEA Safety Standards for an infinite shipment (N=oo) with a zeroCriticality Safety Index.

Table 6.122. Package Parameters for the Model 2000 Cask with TRIGA Case XVIIType of Fuel (70% Enriched)



Fuel Material UZrH1.6 (8.5 wt% U)



Maximum Fuel Enrichment 70.0%

23SU mass per divider position 240 grams

Minimum Fuel Burmup N/A

Burnable Poison Materials* 20 to 36 grams per fuel pin (Not Modelled)


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6.7.8.5 Packagc Fuel Loading

Table 6.150 summarizes the maximum fuel load (mass), as well as some significant constraintsfor the form of the'MTR Type Ill-B fuel. The loading specified in Table 6.150 described hereinqualifies as an infinite (N=oo) Fissile Class III-B package with a zero Criticality Safety Indexunder the provisions of IOCFR71.61 based on the infinite cask analysis. This loading also meetsthe JAEA Safety Standards for an infinite shipment (N=co) with a zero Criticality Safety Index.

Table 6.150. Package Parameters for Model 2000 Cask with MTR Type HI1-B Fuel (UAI.)

Parameters Quantity/Comment

Fuel type UAI/Al

235U mass per plate 20 grams

Number of plates/fuel element 23

Number of fuel elements per divider position 2


235U enrichment 93.0%

U mass per fuel plate 60.0 grams

Al mass (fuel) per fuel plate 159.434 grams

Al density 2.7 g/cc

6442

CM05007Page 5

ATTACHMENT B

Replacement Pages for Reference (2)

NEDO-31581 March 2005

TABLE OF CONTENTS (Continued)

Page

4.4.1.4 Test Procedures ........................... 4-11

4.4.1.5 Results ................................... 4-12

4.4.1.6 Low Temperature Test Results .............. 4-14

4.4.1.7 Ambient Temperature Test Results .......... 4-14

4.4.1.8 High Temperature Test Results ............. 4-17

4.4.1.9 Conclusions ............................... 4-17

4.4.1.10 References ................................ 4-17

4.4.2 Containment for Failed Fuel ......... .. .............. 4-18

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

5.1 DISCUSSION AND RESULTS . . . 5-1

5.2 SOURCE SPECIFICATIONS . . . 5-4

5.2.1 Gamma Sources ................... .................... 5-5

5.2.2 Neutron Source ................ .. ................... 5-6

5.3 MODEL SPECIFICATION . . . 5-6

5.3.1 Description of the Radial and Axial ShieldingConfiguration ...................................... 5-6

5.3.2 Shield Regional Densities .......................... 5-8

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

5.4.1 Fuel Source ......................................... 5-11

5.4.2 Activation Product Sources ......................... 5-14

5.4.3 Bounding Limits ..................................... 5-15

5.5 APPENDIX .................................................... 5-16

5.5.1 RSIC Code Package CCC-137 .......................... 5-17

5.5.2 RIBD/ISOSHLD Input .................................. 5-25

5.5.3 Sample of the Neutron Dose Rate Calculations at theTop and Side Surfaces of the Model 2000 Package .... 5-133

5.5.4 Horizontal Shipment ............... ................. 5-135

5.5.4.1 Scope ..................................... 5-135

5.5.4.2 Assumptions ................................ 5-135

5.5.4.4 Description of Source Inputs in MCNP ...... 5-137

5.5.5 Horizontal Shipment of LVIR SNF and Hardware/RadioisotopeMaterials . . . . . . . . . . . . . . . . . . . . 5-137a

xi


1. GENERAL INFORMATION

1.1 INTRODUCTION

The Model 2000 Radioactive Material Transport Packaging was developed at

Vallecitos Nuclear Center. The primary use of the packaging is to provide

containment, shielding, impact resistance, and thermal resistance for its

contents during normal and hypothetical accident conditions. The packaging is

designed to transport Type B quantities of radioactive materials and fissile

materials in solid form (Class III). It complies with the Nuclear Regulatory

Commission (NRC) regulations contained in the Code of Federal Regulations,

Title 10, Part 71 (lOCFR71). The package is to be shipped in all modes of

transportation, except by air. The number of packages per shipment is

determined in accordance with the package/content Criticality Safety Index.

Calculations, engineering logic, and all related documents which demonstrate

compliance with regulations are presented in subsequent sections of this

report.

GE Quality Assurance Program, QAP-1, controls design, purchased, fabrication,handling, shipping, storing, cleaning, assembly, inspection, testing,operation, maintenance, repair and modification of the packages. The NRC hasapproved QAP-! Under Docket No. 71-0171 upon demonstration that the QA Planmeets the requirements of Subpart H of 10CFR71.

1.2 PACKAGE DESCRIPTION

The Model 2000 transport packaging is illustrated in Figure 1.2.1. The

calculated gross weight of the transport packaging and its contents is

approximately 33,550 lbs. The packaging consists of two main components:

1) Cask. A lead-filled 304 stainless steel weldment, cylindrical in

shape, and measuring 38.5 in. OD by 71.0 in. high. The lead cask

provides containment and shielding for its contents.

2) Overpack. The overpack is in the form of a double-walled right

cylindrical vessel with a toroidal shell at each end. The

overpack is constructed of Type 304 stainless steel and measures

131.5 in. high with a nominal diameter of 72.0 in. Its primary

function is to act as crash shield during impact events and, in

addition, it provides a thermal barrier to the cask.

1-1


Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 1OCFR71.19(e), January 2004,

Request for Revision of Certification Identification Number to 9228-96

10CFR71 ApplicableParagraphs Compliance Reference

71.31 Contents ofapplication

71.31 (a)

71.31(b)

71.31(c)

Description and evaluation, for all approved contents, ofthe Model 2000 package are given in the references. TheQuality Program also is cited within these documents, asapplicable. GENE/VNC Quality Program has been approved bythe NRC under Docket No. 71-0170. The Model 2000 packageis certified by the NRC under Certificate of Compliance(CoC) No. 9228.

Additional authorized contents have been submitted to theNRC for their approval. Approved by the NRC as CoC 9228Revision 0 through Revision 13. See References.

GE complied with the requirements of NUREG/CR-3854 for thefabrication of the Model 2000 packages, Serial Numbers2002 and 2003, the HFIR Fuel Basket & Liner(A), MTR FuelDivider(B), and TSR Fuel Basket(B). Also, the Multi-Functional Rack(C) was designed to the NUREG. The Model2000 package, Serial No. 2001 was fabricated to therequirements of ASME Code, Section 111, Subsections NG andNF. Maintenance activities of all Model 2000 packages andNRC approved hardware are conducted according with theoriginal design requirements.

NEDO-31581; GE DWGS129D4946, 105E9520,105E9521, 101E8718 &101E8719.NEDO-32229; GE DWG105E9523.NEDO-32318; GE DWGS105E9555, 166D8066 &183C8356.NEDO-32408; GE DWGS105E9557 & 105E9560.

NEDO-32229; GE DWG105E9523.NEDO-32318; GE DWGS105E9555, 166D8066 &183C8356.NEDO-32408; GE DWGS105E9557 & 105E9560.

NEDO-31581, Chapter 8; GEDWGS 129D4946, 105E9520,105E9521, 101E8718 &101E8719.(A) NEDO-32229, Chapter 8;GE DWG 105E9523.(C) NEDO-32318, Chapter 8;GE DWGS 105E9555, 166D8066& 183C8356. (B) NEDO-32408,Chapter 8; GE DWGS 105E9557& 105E9560.

I

1-15


IModel 2000 Radioactive Material Transport Package

Demonstration of Compliance with 1OCFR71.19(e), January 2004,


1OCFR71 ApplicableParagraphs Compliance Reference

71.33 Packagedescription

71.33 (a) A steel encased lead shielded shipping cask. The cask is

transported within a double-walled overpack with toroidal

shell impact limiters at each end. The overpack mitigates

the effects of the specified drop conditions and acts as a

heat shield. The overall dimensions of the packaging are

approximately 131.5 inches in height and 72.0 inches in

diameter. The cask is transported in the upright

position. The gross weight of the package isapproximately 33,550 lbs. A detailed description of the

package, meeting the requirements of this paragraph, is

given in the references.

NEDO-31581, Chapter 1; GEDWGS 129D4946, 105E9520,105E9521, 101E8718 &101E8719.NEDO-32229, Chapter 1; GEDWG 105E9523.NEDO-32318, Chapter 1; GEDWGS 105E9555, 166D8066 &183C8356.NEDO-32408, Chapter 1; GEDWGS 105E9557 & 105E9560.

1-16


Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 10CFR71.19(a), January 2004,


1OCFR71 ApplicableParagraphs

71.33(b)

Compliance

The Model 2000 package is approved for transporting thecontents given below. The cited references detail thecontents identification and maximum quantities ofradioactivity of radioactive and fissile constituents, thechemical and physical form, the extent of reflection, theamount and identity of non-fissile materials used asneutron absorbers or moderators, configuration requiredfor nuclear safety, material density, maximum weight,maximum amount of decay heat, loading restrictions, ifapplicable, etc.

Content (i)(A) includes fuel rods, which may be cut orsegmented. These rods are contained within a closed (butnot leak-tight) 5inch Schedule-40 pipe with a maximumusable length of 39-5/8 inches with no more than 437.5 gfissile per pipe. Limits per package are:

Reference

(A) NEDO-31581, Chapters 1,5, & 6; GE DWGS129D4946,105E9520,105E9521, 101E8718 &101E8719.(B) NEDO-32229, Chapters 1,5, & 6; GE DWG 105E9523.(A) NEDO-32318, Chapters 1,5, & 6; GE DWGS 105E9555,166D8066 & 183C8356.(C) NEDO-32408, Chapters 1,5, & 6; GE DWGS 105E9557 &105E9560.

(a) 1175g fissile with U-235 enrichment not exceeding 5w/o and the minimum pellet diameter>= 0.30 inch.

(b) 1750g fissile with U-235 enrichment not exceeding 5w/o and the minimum pellet diameter >= 0.35 inch.

(c) 5OOg of U-235 for pellet diameter < 0.30 inches.

1-17

NEDO-31581 M~arch 2005

Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with l0CFR71.19(e), January 2004,


lOCFR71 ApplicableParagraphs Compliance Reference

71.33(b)(cont.) Content (ii)(A) includes the following:

(a) Byproducts and source materials in solid form with amaximum decay heat of no more than 600 watts. Ashipping limit of 500 grains of U-235 equivalent massof special nuclear material is applicable.

(b) Pu in excess of 20 curies/package in the form ofmetal, metal alloy, reactor fuel elements, or inspecial solid nuclear material, or contained inaccordance with lOCFR71.63(6). The loading shall notexceed 300 grams of Pu-239-equivalent mass.

Content (iii)(B)includes HFIR fuel assembly, whichconsists of two concentric cylindrical elements ofinvolute-shaped U(93%) Al fuel plates clad with Al. Theouter element has 6872 grams of U-235 maximum and theinner element has 2628 grams of U-235 maximum.

Content (iv)(C) includes the TSR fuel loading, whichconsists of 2677 grams of U-235 at the top section, 304grams at the middle section, and 1412 grams at the bottomsection. The maximum fuel enrichment is 94% for all threesections.

1-18


71.33(b) (cont.)

NEDO-31581

Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 1OCFR71.19(a), January 2004,


Compliance

Content (v)(C) includes the following:

(a) 14910 grams of U-235 (94% enrichment) for U308-typefuel,

(b) 14595 grams of U-235 (20% enrichment) for U3Si2-typefuel,

(c) 23352 grams of U-235 (94% enrichment) for UALx-typefuel, and

(d) 4620 grams of U-235 (95% enrichment) for U metal-typefuel.

Content (vi)(C) includes the following:

(a) 28770 grams of TRIGA Case XVII type fuel (20%enrichment),

(b) 28770 grams of TRIGA Case XVII type fuel (70%enrichment),

(c) 25599 grams of TRIGA Case V type fuel (94%enrichment), and

(d) 10374 grams of TRIGA Case Vl1l type fuel (94%enrichment).

March 2005 |

Reference

1-19





71.35 Packageevaluation

71. 35 (a)

71.35(b)

71.35(c)

Compliance to subpart E and F of Part 71 is demonstratedby engineering analysis, testing, or both, as presented inthe references given.

For contents (i), (ii), and (iii) the Criticality SafetyIndex (CSI) for criticality is 100 and for contents (iv),(v), and (vi) the CSI is 0.0. Refer to "71.33(b)" for.description of contents.

The only special required for the transportation offissile contents, is the requirement to vacuum dry thecask cavity in the event of wet loading.

NEDO-31581,4, 5, & 6.NEDO-32229,4, 5, & 6.NEDO-32318,4, 5, & 6.NEDO-32408,4, 5, & 6.

NEDO-31581,NEDO-32229,NEDO-32408,

NEDO-31581,NEDO-32229,NEDO-32408,

Chapters 2, 3,

Chapters 2, 3,

Chapters 2, 3,

Chapters 2, 3,

Chapters 1 & 6.Chapters 1 & 6.Chapters 1 & 6.

Chapters 1 & 7.Chapters 1 & 7.Chapters 1 & 7.

71.37 Qualityassurance

71.37 (a) GE Quality Program (QAP-1) is described in NRC's DocketNo. 71-0170. Quality Program requirements are included inthe References, as applicable. The application of theQuality Program and its supporting documents assures thatthe quality requirements are identified and adhered toduring the design, fabrication, operation, and maintenanceof the Model 2000 package.

NEDO-31581, Chapters 1 & 8,GE DWGS 129D4946,105E9520,105E9521, 101E8718 &101E8719.NEDO-32229, Chapters 1 & 8,GE DWG 105E9523.NEDO-32318, Chapters 1 & 8;GE DWGS 105E9555, 166D8066& 183C8356.NEDO-32408, Chapters 1 & 8;GE DWGS 105E9557 &105E9560.

1-20

NEDO-31581 March 2005I

Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 10CFR71.19(e), January 2004,



71.37(b) All 18 criteria of QAP-l are applicable to the design,fabrication, operation and maintenance of the Model 2000package. Application of these criteria are defined withina series of documents at GENE/VNC. GENE/VNC documents arefurther defined and amplified by GENE Quality AssuranceProgram, NEDO-11209-04 as accepted by NRC under Docket No.71-0380. A series of tests are performed duringfabrication and upon completion of the Model 2000 packageto establish its acceptance: visual and dimensionalinspections of components and welds, NDE examinations ofwelds per ASME Code Section III, leak test of thecontainment boundary prior to lead pouring and of thecask, leak test of cask seal surrogate test-coupon perANSI N14.5, crush test of the honeycomb material,shielding integrity test using a gamma source, hydrostatictest on the cask cavity for structural integrity of thecontainment system, and a thermal test of the assembledpackage is performed for verification of the analyticalmodel.

NEDO-31581,NEDO-32229,NEDO-32318,NEDO-32408,

Chapter 8.Chapter 8.Chapter 8.Chapter 8.

Subpart E-PackageApproval Standards

71.41Demonstration ofcompliance

71.41(a) The effects on the packages of the normal (71.71-) andaccident (71.73) conditions of transport are demonstratedby engineering analysis (Finite Element Method, FEM).testing, or by the application of both, as presented inthe references given.


Chapters 2 & 3.Chapters 2 & 3.Chapters 2 & 3.Chapters 2 & 3.

1-21

NEDO-31581 E 5March 2005




71. 41(b)

Compliance

The Model 2000 package is evaluated independently of itstransportation trailer.

No environmental or test conditions different from thosespecified in 71.71 and 71.73 are evaluated.

Reference





71.41(c)

71.43 Generalstandards for allpackages

71.43 (a) The smallest overall dimension of the Model 2000 isgreater than the minimum dimension given in this paragraphof Part 71. Refer to Paragraph 2.4.1 of the givenreferences.



71.43(b) The Model 2000'package has a feature for the installationof a wire seal across the overpack closure. While intact,the wire seal would be evidence that the package has notbeen opened by unauthorized persons. Refer to Paragraph2.4.2 of the given references.

NEDO-31581, Chapter 2.GE DWGS 105E9521 &101E8719.NEDO-32229, Chapter 2.NEDO-32318, Chapter 2.NEDO-32408, Chapter 2.

71.43(c) The Model 2000 cask is closed by 15 1.25 inches diametersocket head bolts. The bolts are tighten to 690 Ft-Lbs.Refer to Paragraph 2.4.3 of the given references.



1-22


Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 1OCFR71.19(o), January 2004,


1OCFR71 ApplicableParagraphs Compliance

71.43(d) The materials of construction and finishing selection forthe cask, baskets, racks, and dividers assures that nosignificant chemical, galvanic, or other reaction couldoccur in the package. Refer to Paragraph 2.4.4 of thegiven references.

Reference

NEDO-31581, Chapter 2; GEDWGS 129D4946, 105E9520,105E9521, 101E8718 &101E8719. NEDO-32229,Chapter 2; GE DWG 105E9523.NEDO-32318, Chapter 2; GEDWGS 105E9555, 166D8066 &183C8356.NEDO-32408, Chapter 2; GEDWGS

105E9557 & 105E9560.GE DWGS 105E9520 &101E8718.

71.43(e) The Model 2000 packaging does not have a valve or otherdevice that would allow the release of radioactivematerials upon failure.

71.43(f) The Model 2000 package design features, fabricationtechnique, and preparation for shipment assure that undernormal conditions of transport there would be no loss ordispersal of radioactive content, no significant increasein external radiation levels, and no substantial reductionin the effectiveness of the packaging. This is verifiedby the initial drop test and by the annual and pre-shipment inspections.

Table 3.1.1 of the given references show that under thecondition of still air at 380C (1000F) and in the shade noaccessible surface of the package would have a temperatureexceeding 50'C (1220 F).

NEDO-31581; Chapters 2, 47, GE DWGS 105E9520, &101E8718.NEDO-32229; Chapters 2, 4& 7.NEDO-32318; Chapters 2, 4& 7.NEDO-32408; Chapters 2, 4& 7.

71.43(g) NEDO-31581,NEDO-32229,NEDO-32318,NEDO-32408,


71.43(h) The Model 2000 cask has been design to preclude continuousventing from its cavity. Refer to the given references.

GE DWGS 105E9520 &101E8718.

1-231



Demonstration of Compliance with lOCFR71.19(e), January 2004,



71.45 Lifting andtie-down standardsfor all packages

71.45 (a)

71.45(b)

There are three types of lifting devices use in the Model

2000 cask: (1) Standard ears; (2) Auxiliary ears; and (3)

Trunnion. Except for these devices, there are no other

devices or features of the cask that could be used forlifting the package, once the cask is within the overpack.

The engineering evaluation (closed form solution and

finite element analysis) presented in Paragraph 2.5.1 of

the given references, show that all of the lifting device

designs have a minimum safety factor of three againstyielding during lifting operations. Failure of thesedevices will not prevent the package from meeting other

structural requirements.

The Model package has eight tie-down points, referred as

Tie-down ribs. These tie-down ribs are the only points on

the exterior of the package that could be used for tying

the package to its trailer. The engineering evaluation

(closed form solution and finite element analysis)presented in Paragraph 2.5.2 of the given references, show

that the tie-down ribs are capable of withstanding the

required transportation loads without generating stresses

in the material in excess of the material yield stress.Under excessive load, failure of these devices will not

impair the package ability to meet other requirements.

NEDO-31581, Chapter 2.NEDO-32229, Chapter 2.GE DWGS 129D4946,105E9520,& 101E8718.

NEDO-31581, Chapter 2.NEDO-32229, Chapter 2.GE DWGS 129D4946, 105E9521,& 101E8719.

1-24





71.47 Externalradiationstandards for allpackages

71. 47 (a)

71.47(b)

The Model 2000 cask is designed to be transported byexclusive use shipment only. The radiation level limitsof 71.47(a) do not apply to an exclusive use shipment.Refer to Chapter 5 of the given references.

The Model 2000 cask is designed to reduce the radiationfrom the maximum allowed radioactive contents to levelswhich comply with the limits of 71.47(b). Refer to Tables5.1.1; 5.1; 5.1.1 & 5.1.2; and 5.1A, 5.1B, & 5.1C;respectively, of the given references for the summary ofmaximum dose rate obtained by the engineering evaluation.The following computer codes were used in the evaluation:from the Radiation Shielding Information Center (RSIC),RIBD, ISOSHLD, & ORGIN; from Los Alamos NationalLaboratory (LANL) MCNP.





71.47(c) Specific written instructions forincludes controls provided to theshipping papers on every GENE/VNC

exclusive use shipmentcarrier, with theshipment.

Form number SR-170."Expedited Service, andExclusive Use VehicleInstructions", dated 5/97. I

71.47(d) The written instructions for exclusive use shipments aresufficient so that, when followed, they will cause thecarrier to avoid actions that will unnecessarily delaydelivery or unnecessarily result in increased radiationlevels or radiation exposures to transport workers ormembers of the general public.

Form number SR-170,"Expedited Service, andExclusive Use VehicleInstructions", dated 5/97.

1-25

NEDO-31581 March 2005 |




71.51 Additionalrequirements forType B packages

71.51(a)

71.51(b)

The Model 2000 packaging is designed, constructed, andprepared for shipment to meet and maintain the leaktightness criteria given in ANSI N14.5 under normal andaccident conditions of transport. This is demonstrated bythe structural (Chapter 2) and thermal (Chapter 3)analyses and the prototype testing (Chapter 2, Paragraph2.7.5 and Chapter 4, Paragraph 4.4. 1) presented in thereferenced document. In addition, the operational andmaintenance procedures (Chapter 7 and 8). assurescontinuous compliance with the ANSI requirement.

See 71.51(a) above.

NEDO-31581

71.51(c) The leak tightness characteristic of the Model 2000package does not depend on filters or cooling systemdevices.

NEDO-31581,& 4.NEDO-32229,& 4.NEDO-32318,& 4.NEDO-32408,& 4.

Chapters 2, 3,

Chapters 2, 3,

Chapters 2, 3,

Chapters 2, 3,

71.52 Exemptionfor low-specific-activity (LSA)packages

71.53 Fissilematerialexemptions

The Model 2000 package meets the requirements of Paragraph71.51. The Model 2000 is license as Type B package for avariety of contents including fissile materials.

The Model 2000 package meets the requirements of Type Bpackages for a variety of contents including fissilematerials.

NEDO-31581

NEDO-31581

1-26





71.55 Generalrequirements forfissile materialpackages

71.55(a)

71.55(b)

The Model 2000 package has been constructed in compliancewith Paragraph 71.41 through 71.47, and Paragraph 71.51.

The GE MERIT (GEMER) computer code was used in thecriticality evaluation. GEMER is an enhanced version ofthe MERIT Monte Carlo code, incorporating the cross-section processing of MERIT and the geometry handling ofKENO-IV. There are 190 energy groups in the cross-sectionset, with the resolved resonances treated explicitly bykeeping track of each neutron's energy and the Breit-Wigner's equation at each collision. Critical benchmarkswere extensively performed to validate the GEMER code (seeChapter 6.5 of NEDO-32408). The Model 2000 package hasbeen designed and constructed such that its contents wouldbe subcritical even if water were to leak into thecontainment system. In addition, the code bias wasdeveloped-as shown in Section 6.5 of NEDO-32408. For allcontents, the Model 2000 package has been analyzed for themost reactive credible configuration, optimum watermoderation, and full-water reflection (for CSI=100) orleakage-free (for CSI-0) boundary conditions. The k-effectives for these cases (normal, accident, or infinitecask array) are all below the licensing limit of 0.95.





I

1-27


Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with l0CFR71.19(e), January 2004,



71.55(c) The Model 2000 packaging is designed, constructed, andprepared for shipment to meet and maintain the leaktightness criteria given in ANSI N14.5 under normal andaccident conditions of transport. This is demonstrated bythe structural (Chapter 2) and thermal (Chapter 3)analyses and the prototype testing (Chapter 2, Paragraph2.7.5 and Chapter 4, Paragraph 4.4.1) presented in thereferenced documents. In addition, the operational andmaintenance procedures (Chapter 7 and 8) assure continuouscompliance with the leak tightness requirement.

NEDO-31581, Chapter 1.NEDO-32229, Chapter 1.NEDO-32318, Chapter 1.NEDO-32408, Chapter 1.

I

1-28



Request for Revision of Certification Identification Number to 9228-96 I1OCFR71 Applicable

Paragraphs

71.55(d)

Compliance

Under normal conditions of transport, the contents are allshown to be subcritical. The Model 2000 package has beenanalyzed for the most reactive credible configuration,optimum water moderation, and full-water reflection (forCSI=100) or leakage-free (for CSI=0) boundary conditions.For Content (i.a) the criticality evaluations for 1175grams of U-235 equivalent mass are shown in Figure 6.4.2.6in NEDO-31581, April 1988.For Content (i.b), the criticality evaluations for 1750grams of U-235 equivalent mass are shown in Figure 6.4.2.8in NEDO-31581, April 1988.For Content (ii.b) the criticality evaluations for 300grams of Pu-239 equivalent mass are shown in Figure6.4.2.4 in NEDO-31581, April 1988.Contents (i.c) and (ii.a) are enveloped by the criticalityevaluations for 500 grams of U-235 equivalent mass modelin Figure 6.4.2.2 in NEDO-31581, April 1988.For contents (iii), the criticality results for the normalarray cases are summarized in Table 6.4 in NEDO-32229,July 1993.For contents (iv),(v), and (vi), the criticality resultsfor the normal conditions are bounded by the Infinite CaskMethodology cases.For contents (iv), the criticality results for theinfinite array cases are summarized in Tables 6.58 ofNEDO-32408 Rev. lb. June 1997.For contents (v), the criticality results for the infinitearray cases are summarized in Tables 6.8, 6.8, 6.9, and6.10 of NED032408 Rev. lb, June 1997.For contents (vi), the criticality results for theinfinite array cases are summarized in Tables 6.11, 6.12,and 6.13 of NEDO-32408Rev. lb, June 1997.And all k-effectives are shown to be below the licensinglimit of 0.95.

Reference

NEDO-31581, Chapter 6.NEDO-32229, Chapter 6.NEDO-32318, Chapter 6.NEDO-32408, Chapter 6.

I

1-29


Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with lOCFR71.19(e), January 2004,



71.55(e) The hypothetical accident conditions are covered by theInfinite Cask Methodology for contents (iv), (v), and (vi)for the most reactive credible configuration, optimumwater moderation, and leakage-free boundary conditions.The Model 2000 package has also been analyzed as two side-by-side containers for contents (i), (ii), and (iii) forthe most reactive credible configuration, optimum watermoderation, and full-water reflection boundary condition.For Content (i.a), the criticality evaluations for 1175grams of U-235 equivalent mass are shown in Figure 6.4.2For Content (i.b), the criticality evaluations for 1750grams of U-235 equivalent mass are shown in Figure 6.4.2.7in NED031581, April 1988.For Content (ii.b), the criticality evaluations for 300gPu-239 equivalent mass are shown in Figure 6.4.2.3 inNEDO-31581, April 1988.Contents (i.c) and (ii.a) are enveloped by the criticalityevaluations for 5OOg U-235 equivalent mass model in Figure6.4.2.1 in NEDO-31581, April 1988.For contents (iii), the criticality results for theaccident array cases are summarized in Tables 6.4, 6.5,and 6.6 in NEDO-32229, July 1993.For contents (iv),(v), and (vi), the criticality resultsfor the accident conditions are bounded by the infinitecask array cases.For contents (iv), the criticality results for theinfinite array cases are summarized in Tables 6.58 ofNEDO-32408 Rev. lb, June 1997.For contents (v) the criticality results for the infinitearray cases are summarized in Tables 6.8, 6.8, 6.9, and 6.10 of NEDO-32408 Rev. lb, June 1997.For contents (vi), the criticality results for theinfinite array cases are summarized in Tables 6.11, 6.12,and 6.13 of NEDO-32408, June 1997.And all k-effectives are shown to be below the licensinglimit of 0.95.



1-30





71.59 Standardsfor arrays offissile materialpackages

71.59 (a)

71.59(b)

71.59(c)

The Model 2000 package has been analyzed in an infinitecask array configuration for contents (iv), (v), and (vi).Hence the value of N is X for contents (iv), (v), and(vi). In addition, the Model 2000 package has beenanalyzed as a single container or side-by-side containersfor contents (i), (ii), and (iii). Hence a value of N-1/2is justified for contents (i), (ii), and (iii).

The CSI is zero for contents (iv), (v), and (vi) as thevalue of N=c. The CSI is 100 for contents (i), (ii), and(iii) with a N value of 1/2.

The CSI is zero for contents (iv), (v), and (vi). TheModel 2000 package may be shipped by any carrier for fuelswith a CSI of zero. The CSI is 100 for contents (i),(ii), and (iii). For fuels with a CSI of 100, the Model2000 package must be shipped by exclusive use vehicle orother shipper controlled system specified by DOT forfissile material packages. This requirement is compliedwith by shipping the Model 2000 package in a GE authorizedvehicle specified in Chapter 1 of the given references.







I

71.61 Specialrequirement forIrradiated nuclearfuel shipments

The engineering evaluation presented in Appendix 2.10.10,.Table 2.10.10.3, of the given reference show that thecritical hoop stress under hydrostatic pressurecorresponds to 40 Mpa, therefore an external pressure of 2Mpa (290 psi) has negligible effect on the cask body.

NEDO-31581

1-31

NEDO-31581 March 2005 |

Model 2000 Radioactive Material Transport PackageDemonstration of Compliance with 10CFR71.19(a), January 2004,


lOCFR71 ApplicableParagraphs Compliance Reference

Subpart F-Package,Special Form, andLSA-111 Tests

71.71 NormalConditions oftransport

71.71(a) The evaluation of the Model 2000 package for Normalconditions of transport, consists of a series ofstructural and thermal analyses with verification testingof key design features and components. All analyses wereperformed using the finite element method. The analysesfollow the guidelines given in Regulatory Guides 7.8 and7.6. Detailed description of the evaluation is given inSection 2.6 of the references and the results aresummarized in Tables 2.6.11.1, 2.15, 2.9, 2.15, and 2.25,respectively. Tables 2.15 and 2.25 are part of the samereference (NEDO-32408). Differential thermal expansionand lead pouring process effects are accounted in theevaluation.

The structural evaluation for Normal conditions oftransport was based on the ambient temperature of -290C(-20'F) and +38'C (+1000 F). Refer to Section 3.4 of thegiven references.



Chapters 2.Chapters 2.Chapters 2.Chapters 2.


71.71(b)

71.71(c)(1) Paragraph 3.4.1.3 of the given references, demonstrateshow Insolation Data is applied in the engineeringevaluation,

.NEDO-31581, Chapters 3

71.71(c)(2) Paragraph 2.6.2 of the given references, describes theapproach taken to demonstrate compliance with Coldcondition requirements.



1-32

NEDO-31581 M1arch 24005




71.71(c)(3)

71.71(c)(4)

Paragraph 2.6.3 of the given references, describes theapproach taken to demonstrate compliance with Reducedexternal pressure condition requirements.

Paragraph 2.6.4 of the given references, describes theapproach taken to demonstrate compliance with Increasedexternal pressure condition requirements.





71.71(c)(5)

71.71(c)(6)

71.71(c)(7)

Paragraph 2.6.5 of the given references,approach taken to demonstrate. compliancecondition requirements.

Paragraph 2.6.6 of the given references,approach taken to demonstrate compliancecondition requirements.

Paragraph 2.6.7 of the given references,approach taken to demonstrate compliancecondition requirements.

describes thewith Vibration

describes thewith Water spray

describes thewith Free drop







71.71(c)(8) Paragraph 2.6.8 of the given references,approach taken to demonstrate compliancecondition requirements.

describes thewith Corner drop


ChaptersChaptersChaptersChapters

2.2.2.2.

71.71(c)(9) Paragraph 2.6.9.of the given references,approach taken to demonstrate compliancecondition requirements.

describes thewith Compression



1-33





71.71(c)(10) Paragraph 2.6. 10 of the given references, describes theapproach taken to demonstrate compliance with Penetrationcondition requirements.



71.73 Hypotheticalaccidentconditions

71.73 (a)

71.73(b)

The evaluation of the Model 2000 package for Accidentconditions of transport, consists of a series ofstructural and thermal analyses with verification testingof key design features and components. All analyses wereperformed using the finite element method. The analysesfollow the guidelines given in Regulatory Guides 7.8 and7.6. Detailed description of the evaluation is given inSection 2.7 of the references and the results aresummarized in Tables 2.7.6.1, 2.20, 2.11, 2.20, and 2.26,respectively. Tables 2.20 and 2.26 are part of the samereference (NEDO-32408).

The structural evaluation for Accident conditions oftransport was based on the ambient temperature of -290C(-200F) and +380C (+100'F). Refer to Section 3.5 of thegiven references.

Paragraph 2.7.1 of the given references, describes theapproach taken to demonstrate compliance with the Free.drop requirements.

Paragraph 2.7.2 of the given references, describes theapproach taken to demonstrate compliance with the Puncturerequirements.





71.73(c)(1)

71.73(c)(3)





1-34





71.73(c)(4)

71.73(c) (5)

71.73(c)(6)

Paragraph 2.7.3 of the given references,approach taken to demonstrate compliancerequirements.

Paragraph 2.7.4 of the given references,approach taken to demonstrate complianceImmersion-fissile material requirements.

Paragraph 2.7.5 of the given references,approach taken to demonstrate complianceImmersion-all packages requirements.

describes thewith the Thermal

describes thewith the

describes thewith the







Subpart G-Operating Controlsand Procedures

71.81Applicability ofoperating controlsand procedures

71.83 Assumptionsas to unknownproperties

The Model 2000 package operations and maintenancerequirements are given in the reference documents. Theseactivities are conducted within the requirements frameworkof GENE/VNC quality program, QAP-1, Docket No. 71-0170.

GENE has the computational capability of developing orobtaining all material properties pertinent to shieldingand criticality evaluation. Some of the properties aregiven as limiting case, e.g. source term for shielding andfresh fuel for criticality.


NEDO-31581,6.NEDO-32229,6.NEDO-32318,6.NEDO-32408,6.

Chapters 7Chapters 7Chapters 7Chapters 7

& 8.& 8.& 8.& 8.

IChapters 1, 5 &

Chapters 1, 5 &

Chapters 1, 5 &

Chapters 1, 5 &

1-35

NEDO-31581



March 2005 |


71.85 Preliminarydeterminations

71.85 (a) A series of tests are performed during fabrication andupon completion of the Model 2000 package to establish itsacceptance: visual and dimensional inspections ofcomponents and welds and NDE examinations of welds perASME Code Section 111, Subsections NB and NG, are amongthe inspections performed. NDE inspection requirementsare detailed, for each weld, in the referenced drawings.

The Model 2000 cask cavity is hydrostatically tested perParagraph NB6200 of the ASME Code, Section III, SubsectionNB. The test pressure is 1.5 the design pressure (45psi).

The name plates attached to the cask and overpack of theModel 2000 package are shown in the given references.

NEDO-31581, Chapter 8; GEDWGS 129D4946, 105E9520,105E9521, 101E8718 &101E8719.NEDO-32229, Chapter 8; GEDWG 105E9523.NEDO-32318, Chapter 8; GEDWGS 105E9555, 166D8066 &183C8356.NEDO-32408, Chapter 8; GEDWGS 105E9557 & 105E9560.

NEDO-31581, Chapters 8.

GE DWGS 129D4946,105E9520,105E9521, 101E8718 &101E8719.

71.85(b)

71.85(c)

71.87 Routinedeterminations

Engineering evaluation is performed of the proposedcontents prior to each shipment of the Model 2000 package,to assure that the packaging and its contents satisfiesthe requirements and limitations imposed by theCertificate of Compliance (CofC). The evaluation scopeincludes a review of thermal, criticality, shielding, andmechanical aspects of the proposed contents. In additionthe an evaluation of operational aspects is also performedto assure that the packaging can be effectively utilizedfor the proposed contents.

NEDO-31581, Chapters 7.NEDO-32229, Chapters 7.NEDO-32318, Chapters 7.NEDO-32408, Chapters 7.

1-36





71.89 Openinginstructions

GENE/VNC issues Operations and Maintenance procedure toall users of the Model 2000. The procedure limits theperformance of all major repair activities on safetyrelated components to GENE/VNC personnel only.


Chapters 7. |Chapters 7.Chapters 7.Chapters 7. |

71.91 Records

71.93 Inspectionand tests

71.95 Reports

Quality records of each package are kept at GENE/VNC'sVallecitos Nuclear Center in Pleasanton, California.

The NRC has conducted several audits of the GENE/VNCQuality Program and the Program application to Type Bpackages.

Over the years of operation of the Model 2000 package,GENE/VNC has kept the NRC informed of events that may havereduced the effectiveness of the package, of defectsdiscovered, or of instances non compliance.

I

Subpart H-QualityAssurance

71.101 Qualityassurancerequirements

71.101 (a) The Quality Program assures compliance with therequirements of Subpart H. The program assures therequired management efforts, equipment and procedures aredirected toward satisfying the quality objectives of GENEof providing safe and reliable systems and components, andassuring compliance with the applicable codes andregulations of governing regulatory agencies.

QAP-1, Docket No. 71-0170

1-37

NEDO-31581 Moarch 2005


Demonstration of Compliance with IOCFR71.19(e), January 2004,Request for Revision of Certification Identification Number to 9228-96

lOCFR71 ApplicableParagraphs. Compliance

71.101(b)

71.101(c)

All 18 criteria of QAP-1 are applicable to the design,fabrication, operation and maintenance of the Model 2000package. Application of these criteria are d6fined withina series of documents at GENE/VNC. GENE/VNC documents arcfurther defined and amplified by GENE Quality AssuranceProgram. NEDO-11209-04 as accepted by NRC under DocketNo. 71-0380.

Both QAP-l and NEDO-11209-04 Quality programs have beenapproved by the NRC under Docket Nos. 71-0170 and 71-0380respectively.

Reference

QAP-1, Docket No. 71-0170NEDO-11209-04, Docket No.71-0380.

QAP-1, Docket No. 71-0170NEDO-11209-04, Docket No.71-0380.

I

71.101(f) NEDO-11209-04 Quality program has been previously approved NEDO-11209-04, Docket No.by the NRC under 1OCFR50, Appendix B.

1-38

NEDO-31581 March 2005 |

1.3.2 Drawings

129D4946,

105E9520,

105E9521,

101E8718,

101E8719,

129D4922,

Rev 10

Rev 4

Rev 5

Rev 12

Rev 12

Rev 2

Model

Model

Model

Model

Model

Model

2000

2000

2000

2000

2000

2000

Transport Container

Shipping Cask All S/N's except S/N 2001

Cask Overpack All S/N's except S/N 2001

Shipping Cask S/N 2001

Cask Over Pack S/N 2001

Cask Liner Certified Drawing

1.4 References

1.4.1 Code of Federal Regulation, Title 10, Part 71, 2004

1.4.2 NEDO-32220, "Model 2000 Radioactive Material Transport Package, HFIR

Fuel Basket and Liner Safety Analysis Report," July 1993

14.3 NEDO-32318, "Model 2000 Radioactive Material Transport Package, 2000

Watts Decay Heat Upgrade Safety Analysis Report," July 1994

14.4 NEDO-32408 Model 2000 Radioactive Material Transport Package,MTR-Type

Fuel Divider and Tower Shielding Reactor Fuel Basket Safety Analysis

Report," December 1994

1-39

CM05007Page 6

ATTACHMENT C

Appendix 5.5.5, Reference (2)

NEDO- 31581 March 2005

Appendix 5.5.5 Horizontal Shipment of the Model 2000 Transportation Package with LWRSNF and 2000 Watt Decay Heat Source - Shielding Analysis

1.0 PURPOSE

The purpose of this calculation is to determine the dose rates for the Model 2000 package shipped in ahorizontal configuration.

2.0 METHODOLOGY

The methodology of Section 5.5.4 of the Model 2000 Safety Analysis Report (SAR) [5.1] for calculatingdose rates is to be the basis for this analysis. The Model 2000 package needs to be shipped in a horizontalposition under certain conditions. As a result of tilting the package to a horizontal position, the originalbottom of the package becomes the back and the original top of the cask becomes the front of thepackage. Figure 1 shows the horizontal shipping configuration.

As stated in Section 5.5.4.2 of the Model 2000 SAR [5.1], the closest approachable surface at the back is89.5 inches from the back overpack. No credit is taken for any of the structures between the backoverpack and the outside surface. The distance to the cab of the truck from the front over pack is 254.23inches (6.45 m). The 2 m dose rate for exclusive use is measured from the back end of the truck and isthus 5.93 m from the closest approachable surface a the back. Figure 1 shows the distance from the frontof the package (the cradle) to the cab to be 6.4 m. In the cab dose rate analysis, the distance from the frontoverpack to the cab was taken to be 6.4 m, thus making this calculation conservative (credit was not takenfor an additional 13.75').

Dose rates are calculated for two types of sources: Light Water Reactor (LWR) spent nuclear fuel (SNF),and a Co-60 source with 2000 Watts of decay heat. The source for the LWR SNF is the same source aswas used in the shielding evaluation of the Cask 2000 SAR [5.1], and the Co6O source is the same sourceas was used in the shielding evaluation of the Model 2000 Decay Heat Upgrade SAR [5.2]. The Co-60source was selected as the bounding 2000 Watt decay heat source as a result of its high energy photons(1.173 and 1.332 MeV), which penetrate the cask body more easily. Dose rates were calculated using theGE Level-2 ECP MCNPOIA.

All dose rates from Section 5.1 of the Model 2000 SAR [5.1] for the LWR SNF and fromSection 5.1 of the are Model 2000 Decay Heat Upgrade SAR [5.2] for the Co-60 source are validand bounding with the exception of the normal condition original bottom (now back) and the cab(now front) dose rates.

As was done in Section 5.5.4.7 of the Model 2000 SAR [5.1], the dose rates of interest are at the backsurface of the package (labeled as "back dose" in Figure 1), 2 meters from the back surface of the truck,and at the cab of the vehicle from the front of the cask.

5.137a

NEDO - 31581 March 2005

bDaom dooo'5KMI7

Figure I - Horizontal Shipment Normal Dose Rate Locations (dimensions in meters and inches)

3.0 INPUTS/ASSUMIPTIONS

3.1 LWVR SINT Source

3.1.1 Gamma Source Term

The gamma source term for the LWR SNF is identical to the gamma source specified in Section 5.2.1 ofthe Model 2000 SAR [5.1]. Table 5.2.1.1 of the SAR [5.1] is recreated below.

Table 1 -LWR Fuel Gamma Source Term

Group Total Group Production Group AverageEnergy (NeV)

Rate (photons/sec)1 4.077E-10 1.500E-022 1.483E+12 2.500E-023 4.777E+1 1 3.500E-024 6.210E+12 4.500E-025 1.617E+10 5.500E-026 9.120E+08 6.500E-027 8.036E+05 7.500E-028 2.003E+13 8.500E-029 9.592E+10 9.500E-02

10 1.606E+14 1.500E-0111 9.590E+12 2.500E-01

5.137b

NEDO - 31581 March 2005

12 8.194E+ll 3.500E-OI13 2.651E+14 4.750E-0114 6.571E+14 6.500E-0115 1.030E+15 8.250E-0116 L029E+13 1.OOOE+0017 5.039E+12 1.225E+0018 8.363E+12 1.475E+0019 6.630E+10 1.700E+0020 2.127E+10 1.900E+0021 6.485E+12 2.100E+0022 5.430E+1 1 2.300E+0023 7.741 E+10 2.500E+0024 2.055E+00 2.700E+0025 2.838E+09 3.OOOE+00

TOTAL 2.183E+15

3.1.2 Neutron Source Term

The neutron source term is 8.75E+07 neutrons/see, as stated in Section 5.2.2 of the SAR [5.1]. The Cf-252 neutron energy spectrum is used in the MCNPOIA runs.

3.1.3 MCNPOIA Model Description

The geometry of the cask and overpack is described in the Reference [5.3] and [5.4] drawings. Credit wastaken for the optional lead-filled liner in Section 5.1 of the Model 2000 SAR [5.1] to provide additionalphoton shielding; this lead-filled liner [5.5] is also used in this analysis. The key cask and overpackdimensions are provided below.

CaskInside Height . 137.16 cm (54')Inside Radius =33.655 cm (13.25")Cavity Shell .2.54 cm (1")Lead shield on side . 10.16 cm (4")Stainless steel at bottom - 15.24 cm (6")Outside diameter .97.79 cm (38.5")Cask Lid (lead) .14.2748 cm (5.62")Cask lid bottom shell .3.81 cm (1.5')Cask Lid top shell .3.81 cm (1.5')Cask Lid side shell .2.54 cm (l")OverpackInner shell side .1.27 cm (0.5')Thickness of support plate .1.27 cm (0.5')Air gap between shells (bottom) .10.16 cm (4")Air gap between shells (side) .7.62 cm (3")Outer shell is 1.27 cm (0.5') on the top and bottom and 1.27 cm (0.5") on the side

5.137c

NEDO - 31581 March 2005

Top honeycomb .15.24 cm (6")Top torus =60.96 cm (24")Bottom torus =60.96 cm (24")

All stainless steel was assumed to have a density of 7.8 g/cm3 and a composition of 72% by weight ofiron, 2% manganese, 18% chromium, and 8% nickel. The density of lead was taken as 11.34 glcm3. Airwas taken to be at a density of 0.001293 g/cm3.

The source geometry for the LWVR SNF source is identical to the source as described in Section 5.2 of theModel 2000 SAR [5.1]. It is a 3.73" diameter, 45" long cylindrical source consisting of uranium oxideand zirconium. The source density was conservatively chosen to be 3 g/cm3.

The flux to dose conversion was done using the ANSI!ANS-6.1.1-1977 standard [5.9].

Gamma and neutron dose rates for the front (at the cab) are found in the files"Gamma Front.out" and "Neutron Front.out", respectively. The gamma and neutron dose rates for theback are found in the files "Gamma Back.out" and "Neutron Back.out", respectively.

3.2 Co-60 Source

3.2.1 Gamma Source Term

The gamma source term for the Co-60 source is identical to the gamma source specified in Table 5.2.1 ofthe Model 2000 Decay Heat Upgrade SAR [5.2], the applicable portion of which is provided below.

Table 2 -Co-60 Gamma Source Term

Isotope Group Production Rate Group Average(photos/sec) Energy (MeV)

Co-60 4.81 E+15 1.1734.81 E+15 1.332

No neutron source is applicable for the Co-60 source.

3.2.2 MCNPO1A Model DescriptionThe geometry of the cask and overpack is described in the Reference [5.3] and [5.4] drawings. Asdescribed in Table 5.1.2 of the Model 2000 Decay Heat Upgrade SAR [5.2], the shipping configurationof this source consists of the barrel rack [5.6] with a 7" minimum wall thickness and the material basket[5.7] with a 1.25 " minimum tungsten plug thickness. The barrel rack and material basket were modeledin this analysis. Due to the nature of the high-energy photons from Co-60, credit was taken for the lead-filled liner [5.5]. The key cask and overpack dimensions, as well as material densities and doseconversion factors, are provided in Section 3.1.3 of this analysis.

The source geometry for the Co-60 source is identical to the source in Table 5.1.2 of the Model 2000

5.137d


Decay Heat Upgrade SAR [5.2]. It is a 9/16" diameter cylinder made of nickel, with a density of 8.9g/cm3.

4.0 Analysis / Results

Tables 3 and 4 below present the results of the dose rates for the locations of interest. All other normalcondition and all accident condition dose rates for the LWR SNF and Co-60 sources are bounded by thedose rates in Section 5 of the Model 2000 SAR [5.1] and Model 2000 Decay Heat Upgrade SAR [5.2],respectively. All dose rates are presented as the calculated dose rate +

*h.1^1 tQilfl-frl nf T IUD f RNFiA& T.r tjnc-. Dntf

Normal Back Surface 2 Meters from the Cab of VehicleConditions of Package Back Surface of Truck from Front

(mremlhr) (mremhr) (mremnhr)Ganuna 10.3 1.29 5.8E.03**Neutron 2.0 0.22 0.24Total 12.3 1.51 0.2549CFRPart 200 10 2173.441

LimitNote: Value was taKen from output file at npsl35,000,000- . . . .

Table 4- Summary of Co-60 Source Dose RatesNormal Back Surface 2 Meters from the Cab of Vehicle

Conditions of Package Back Surface of Truck from Front(mreinhr) (mremlhr) (mremnhr)

Total 48.2 5.68 9.4E-0349CFRPart 200 10 2173.441 1 1 1

It is concluded that all dose rates are within the prescribed limits for both the vertical and horizontalshipments under normal or accident conditions.

5.0 REFERENCES

5.1 NEDO-3 1581, "Model 2000 Radioactive Material Transport Package Safety Analysis Report,"Revision 1, October 2000.

5.2 NEDO-32318, "Model 2000 Radioactive Material Transport Package 2000 Watts Decay HeatUpgrade Safety Analysis Report," Revision 1, August 2000.

5.3 GE Nuclear Energy Drawing No. 105E9520, "Model 2000 Shipping Cask-All S/N's Except S/N2001," Revision 4, General Electric Company, VNC, Pleasanton, CA.

5.4 GE Nuclear Energy Drawing No. 105E9521, "Model 2000 Cask Overpack -All S/N's Except S/N2001," Revision 5, General Electric Company, VNC, Pleasanton, CA.

5.5 GE Nuclear Energy Drawing No. 129D4922, "Model 2000 Cask Liner Certification Drawing,"Revision 2, General Electric Company, VNC, Pleasanton, CA.

5.6 GE Nuclear Energy Drawing No. 166D8066, "Barrel Rack Certification Drawing," Revision 2,General Electric Company, VNC, Pleasanton, CA.

5.137e

NEDO - 31581 March 2005

5.7 GE Nuclear Energy Drawing No. 1 83C8356, "Material Basket Certification Drawving," Revision 2,General Electric Company, VNC, Pleasanton, CA.

5.8 Not used.5.9 ANSI/ANS-6. 1.1-1977, "Neutron and Gamma-Ray Flux-to-Dose-Rate Factors."

5.137f

Documents

GE Nuclear Energy Vallecitos Nuclear Center Sunol, CA ... · [1.2] Model 2000 Radioactive Material Transport Package 2000 Watts Decay Heat Upgrade Safety Analysis Report, NEDO-32318,