ML20211P905

Redacted/non-proprietary Rev 4 to Transtor BWR Criticality Analysis

ML20211P905
Person / Time
Site:	07109268
Issue date:	08/07/1999
From:	Hopf J SIERRA NUCLEAR, INC.
To:
Shared Package
ML20138D213	List: ML20211P835 ML20211P850 ML20211P882 ML20211P905
References
BNFL1.10.06.15, BNFL1.10.06.15-R04, NUDOCS 9909140132
Download: ML20211P905 (40)

Text

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SNC NO.: BFNL01.10.06.15 BNFL-01 CLIENT NO.: _

REVISION NO.: 4 1

DESIGN CALCULATION TranStor BWR Criticality Analysis NOP Prepared by SIERRA NUCLEAR CORPORATION for BNFL

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SNC Sierra Nuclear Corporation Table Of Contents Page

1. of Pages 1

1.0 Introduction / Pumose 3

2.0 Results 11 3.0 Design Input / Assumptions 19 4.0 Methodology / Model 27 5.0 References 28 55 6.0 Calculations Appendix A SCALE 4.1-PC PICTURE Output A1 7

B1 5

Appendix B KENO-Va Input Description C1 4

Appendix C BORAL Sheet Data D1 5

Appendix D Sample KENO-Va Analysis Inputs Appendix E Disk Containing AllInputs / Outputs El 2

Total Number of Pages in Document: 78 panes Revis6on Prepared Date Checked Date Sheet Client / Project BNFL i

4 JEH 8/4/99 KDW g/ggj W

TmnStor" BWR Criticanty Analysis

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1. Intr ~lurrion / Purnose The purpose of this calculation is to establish maximum allowable BWR assembly initial enrichment levels for loading into the BFS TranStor BWR basket. These calculations will verify that all regulatory criticality requirements are met at these maximum enrichment levels. The TranStor" BWR basket is designed to hold a maximum of 61 BWR assemblies during transport and storage. The TranStor* BWR System is designed not to exceed a km f 0.95 (10 CFR71) under worst case cask conditions wh o

is loaded with maximum allowed enriched fuel assemblies. For US BWR fu this calculation determines the maximum allowed enrichment in the 52,60 and 61 assembly TranStor* BWR basket configurations.

Four of the fuel sleeves in the basket have been analyzed to hold canistered fuel debris, or canistered damaged fuel assemblies. The four fuel sleeves lie at the foufcomers of the basket (i.e., the sleeves at the basket edge along the 45 degree angle; the corners of th large 7x7 array of sleeves). These canisters may contain a fuel material configuratio is more reactive than the intact assemblies occupying the other sleeves in the basket without significantly affecting overall basket reactivity, due to their small number and their location at the edge of the basket. Damaged fuel assemblies are assemblies containing one or more fuel rods with significant or visible physical damage. Fuel assemblies with pinhole leaks or hairline cracks are considered intact assemblies. These assemblies do not require confinement canisters and may be loaded into any fuel sleeve within the basket (at the maximum enrichment limits determined for intact fu Analyses have also been performed for partial assemblies occupying the four bas comer sleeves described above, as well as eight additional fuel sleeves that lie along t basket edge. The analyses show that loading partial assemblies around the baske has no significant effect on overall basket reactivity.

SCAM PC; Version 4.3 (Ref.1), a standardized nuclear analysis software system developed for NRC spent fuel storage and transport cask evaluation, will be theTranStor criticality analyzes.

This revised calculation file is written in accordance with QA procedures QAP 3.0, R

~ 7, and QAP 3.2, Rev. 5. Revised QA procedures were in effect by the date this document. However, as documented in Memorandum 99-241 (from Ken Wrig Ram Srinivasan to Howard Wong / Ed Fuller), a request to have this document wr the previous QA procedures was granted.

Revtston T.W Dale CW Date Steet Client / Project BNFL

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4 JEH 8/4/99 KDW 8/6/99 I

W TranStor* BWR Criticality Analysis

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SNC Sierra Nuclear Corporation The following changes have i>een made to the analyses in this revision of the document:

The calculations were modified to use bounding values for the fuel sleeve wall thickness, fuel sleeve spacer thickness, and poison sheet thickness. The nominal fuel sleeve spacer thickness value was also changed (reduced) to reflect the latest design drawings.

The calculations were modified to define a finite list of assembly classes. For each class, all physical parameters important to criticality are specified. Either fixed values or allowable ranges are specified for each parameter. The criticality analyses are performed based upon the allowable values that yield maximum reactivity. For parameters where it is obvious which end of the allowable range yields maximum reactivity, the most reactive value is simply assumed in the analyses. For parameters where it is not obvious, sensitivity studies are performed to determine the most reactive values within the allowable range, The criticality analyses, performed for each assembl class, determine maximum allowable enrichment levels for assemblies that qualify fo that class. Analyses are performed for both the fully and partially loaded BWR baskets.

Criticality analyses for a 52 element BWR basket configuration are added to the calculation. These analyses determine (very high) maximum allowable enrichment levels, by BWR assembly class, for the 52 element configuration.

Additional sensitivity studies are added, including a pellet diameter study, a cladding outer diameter study, a cladding thickness study, and a study of the number and loc of water holes within the assembly rod array.

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analyses with optimum partial assembly arrays replacing intact assemblies in sleeves that lie around the basket edge is performed. This run shows that the 12 assembly arrays will cause no significant increase in overall basket reactivity.

AAer the most reactive damaged assembly rod array is determined, a basket c analysis with optimum damaged assembly arrays replacing intact asse sleeves that lie s'. homers of the basket is performed. This run shows that the 4 damaged assembly array.: will cause no significant increase in overall bask Rohkm Pmpamd Deh Checked Daw Shut C8ent/ Project BNFL 4

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SNC Sierra Nuclear Corporation

2. Results The assembly class specific enrichment limits for the 61,60, and 52 assembly TranStor BWR baskets are provided in Tables 2.1 through 2.3. The 60 assembly basket is identical to the 61 assembly basket except that the central fuel sleeve is left empty. In the 52 assembly basket, the central nine fuel sleeves (a 3x3 array) are left empty.

In addition to the enrichment limits, Tables 2.1 through 2.3 present the results of the criticality analyses which show that the regulatory criticality requirements are met for each assembly class, at the conesponding maximum enrichment level. The analyses demonstrate regulatory compliance in accordance with the methodology specified in NUREG/CR-6361 (Ref. 2). A limiting upper sub-critical limit (USL) is calculated for each case, based upon the average fission group (AFG) parameter (which always yields the limiting USL value). The calculated kerrvalue, plus two times the level of statistical ermr, is shown to be under the limiting USL value for all cases. This verifies compliance with the criticality requirements. In essence, this methodology demonstrates a hg value that is under the regulatory limit of 0.95, after accounting for all code bias and uncertainty effects.

For each case, Tables 2.1 through 2.3 present the AFG value, the limiting USL value that is calculated from the AFG value (using the formulas shown in Section 6 of this value calculated by the criticality code, the statistical error level calculation), the ha (which is output by the criticality code), the final ha value (the calculated value plus value and the times the error level), and finally, the level of margin between the final ha limiting USL value. Thus, all the data that is used to determine the margin between and the limiting USL value (i.e., all the data required to verify compliance with finalLa regulatory requirements), 'are presented in the tables.

A set of BWR assembly classes are defined and presented in Table 2.4. For eac assembly class, Table 2.4 specifies values or allowable value ranges for ever parameter important_to criticality. For the parameters with specified range analyses are performed using the value within the range that yields maximu l

(i.e., the bounding value). In almost all cases, this is the numerical value show 2.4 (not considering the greater or less than sign). This ensures that the criticality analyses which determine the maximum allowable enrichment for each assem are bounding for any assembly that qualifies for that assembly class.

Many analyses are presented in Section 6 of this calculation which verify that,th criticality analysis results presented in Tables 2.1 thmugh 2.3 are bounding fbr.a assemblies that meet the assembly class specifications in Table 2.4. The analyses demonstrate that the maximum enrichment analyses are performed assumjng the mo reactive basket configuration and basket component dimensions (considefing all Revtsion Fm, cd te Checked oste sheet M y oct BNFL 4

JEH 8/4/99 KDW 8/6/99 3

Subject TranStor" BWR Criticality Analysis of 55 i

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E SNC Sierra Nuclear Corporation tolerances). For assembly parameters that have specified ranges in Table 2.4 (as opposed to specific defined values), analyses presented in Section 6 determine what the most reactive values within the specified range are. These bounding values are then used in the maximum enrichment analyses whose results are presented in Tables 2.1 through 2.3.

To qualify for a given assembly class, a specific BWR assembly must meet all of the specifications shown in Table 2.4 for that class. If an assembly qualifies for a given assembly class, the maximum enrichment levels shown for that class in Tables 2.1 through 2.3 are applicable for that assembly. The sets of parameters defined for each class have been carefully selected so that they cover virtually all specific BWR assembly types in the US.

In fact, for some of the parameters which have allowable ranges specified in Table 2.4 (as opposed to specific values), the allowable ranges are extended past the ac~tual BWR assembly values by a small margin. This allows the defined assembly classes to " catch" any assembly types that may be slightly different from the currently known assembly types. It also accounts for tolerances in these parameter values. As a specific example, the minimum allowable values shown for cladding O.D. and cladding thickness are actually slightly lower than the actual nominal values of the BWR assembly types that the defined assembly classes are intended to correspond to. This provides a margin of safety, which allows for tolerances in these parameters, as well as possibly accommodating a variation of that BWR assembly type that has slightly thinner (or smaller) cladding.

The maximum enrichment levels shown for each assembly class in Tables 2.1 through 2.3 are applicable for partial or damaged assemblies as well. Thus, if a given BWR assembly type or design qualifies for a given assembly class (in its intact, design basis state), then a damaged or partial assembly of that assembly type also qualified under that assembly class, with the same maximum allowable enrichment. He enrichment limit for fuel debris is also defined by the assembly type that the debris originated from, ne following restrictions apply, however, for partial and damaged assemblies, as well as fuel debris.

Partial assemblies are restricted to 12 fuel sleeves that lie along the edge of the TranStor BWR basket. This includes the four corner locations of the basket (at the basket edge along the 45 degree axis, i.e., the comers of the 7 x 7 fuel sleeve array). There are rows of three fuel sleeves at the basket edge on each of the four sides of the basket. The two fuel sleeves on the ends of each row (i.e., all but the center sleeve in each row) may also contain partial assemblies. %ese eight locations, along with the four comeriocations, sum to a total of 12 locations. Partial assemblies may be canistered before loading into the basket. Since the TranStor basket designs creates canisters by simply adding screens to the existing fuel sleeves, no modification to the criticality model is necessary to model Revision r.

c cate checked oste sheet W

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4 JEH 8/4/99 KDW 8/6/99 4

Subject TranStor" BWR Criticality Analysis c' E 5

• AM Calculation Number GNFL1.10.OG.1fi

rz SNC Sierra Nuclear Corporation canisters. Thus, from a criticality perspective, both canistered and uncanistered partial

- fuel assemblies may be loaded into the TranStor BWR basket.

Damaged fuel assemblies, or fuel debris, are restricted to the four basket comer fuel sleeves described above. Thus, only four such canisters may be loaded into the basket.

These objects will be canistered. However, as discussed above, the canistering of the damaged fuel has no impact on the criticality models. As discussed in Section 1, damaged assemblies are defined as assemblics with fuel rods that have physical damag greater than pin holes or hairline cracks. Assemblies with pinhole leaks or hairline are considered intact. They do not require a confinement canister and they may be loade into any fuel sleeve in the basket.

Analyses presented in Section 6 of this calculation show that loading 12 partial assemblies, or 4 damaged fuel assemblies around the edge of the basket,has no significan effect on overall basket reactivity. 'Iherefore, the analyses presented in Tables 2.1 through 2.3 apply for partial and damaged fuel assemblies, given the loading restr discussed above. If four damaged assemblies are loaded into the four basket comer sleeves, partial assemblies may be loaded into the eight remaining basket edge sleev (which lie along the side edges of the basket, as described above).

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Sierra Nuclear Corporation S. References

1) " SCALE-4.3: Modular Code System for Performing Computer Analyses for Licensing Evaluations for Workstations and Personal Computers", Oak Ridge National Laboratory, CC-545.

2) NUREG/CR-6361," Criticality Benchmark Guide for Light Water Reactor Fuel in Transportation and Storage Packages", March 1997.

3) BNFL Document BNFLI.10.06.43,"TranStor BWR Basket 30 ft. Drop Analysis.",

Revision 2.

4) BNFL Drawing TSB-010, Sheet 1 of 1, Rev. 4.

5) BNFL Document BNFLI.10.06.14,"TranStor* PWR Criticality Analysis., Rev. 6.

~

6) BNFL Document BNFLI.10.06.50," Shielding Analysis for the TranStor Shipping Package.", Rev. 3.

7) " Characteristics of Spent Fuel, High-Level Waste, And Other Radioactive Wastes Which May Require Long-Term Isolation Volume 3 of 6"; US DOE; DOE /RW-0184; December 1987.

8) BNFL Drawing TSB-002, Sheets 1 and 2 of 2, Rev. 4.

9) BNFL Drawing CA-003, Sheet 1 of 1, Rev. 3.

10) BNFL Drawing CA-002, Sheet 1 of 1, Rev. 3.

I1)BNFL Drawing CA-001, Sheets 2 and 3 of 3, Rev. 3.

12) BNFL Document BNFL 1.10.06.72, " Calculation of Upper Suberitical Limit (USL)",

Rev. O.

13)BNFL Drawing TSB-006, Rev. 2.

14)BNFL Document BNFLI.10.06.16," Benchmark Calculation and Validation of the SCALE 4.1 Code Package (PC Version) For Analysis ofCask Systems with Highly Enriched Fresh Fuel, Fixed Poisons, and Unborated Water", Rev.1.

15) BNFL Drawings TfCl-002 through TfCl-004, Rev.1.

16) BNFL Drawings TTC2-002 through TTC2-004, Rev.1.

17)"BORAL - The Neutron Absorber"; Product Performance Report 624; Brooks &

Perkins Advanced Stmetures.

18)"Nuclides and Isotopes: Chart of the Nuclides",15'h Edition. OE Nuclear Energy, 1996.

4

19) BNFL Document CCV-1.1.4.3, " SCALE Version 4.3 Verification Report.", Rev. O.

Revision Prepared Date Cheded Date sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 27 Subject TranStor" BWR Criticality Analysis of 55 mwe som8 ~ Calculation fiumber St4FL1.10.00.15 l

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JEH 8/4/99 KDW 8/6/99 A-S Subject TranStor"BWR Criticality Analysis of A-7

• A Catcuuon Numbu bNFLUO.0 m

SNC Sierra Nuclear Corporation SLICE AT CENTER OF CASK - 1 SLEEVE - MATEP.IAL MIXTURE MAP MIX 1URE O 1 2 3 4 5 6 7 8 10 SYMBOL 1 2 3 4

5 6 7 8 9 UPPER LEFT LOWER RIGHT COORDINATES COORDINATES X

-1.0000E+01 1.0000E+01 Y

1.0000E+01

-1.0000E401 Z

2.0420E+02 2.0420E+02 U AXIS V AXIS (DOWN)

(ACROSS)

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96 NV=

120 DELU= 2.0833E-01 DELV= 1.5667E-01 J

I l

F l

Revtskm Prepared Date Ched,ed Date Steet Client / Project BNFL 8[4/99 KDW 8/6/99 A-6 4

JEH W

Transtor"BWRCriticality Analysis of A-7 w w.svcp.3 egw u,2, Iwm'a1 C:JrL1.10.0.15

SNC Sierra Nuclear Corporation

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1 i

Revtsion Prepared Date Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 A-7 Md TranStor"BWR Criticality Analysis ny g,7 ww c "'

d u w.al W u C W,W a r-I I

SNC I

Sierra Nuclear Corporation APPENDIX B KENO-Va INPUT DESCRIPTION e

l Checked Date S t Revtskm Prepared pate Qiont/ Project BNFL 4

JEH 8/4/99 KDW 8/6/99 B-1 M eet Transtor BWR Criticality Analysis c ' t.".

"" *

• loams: Calculabon Number BNFLt.10.00.1S I

l l

F SNC Sierra Nuclear Corporation Unit 1 Fuel Pin Cell Fuel Pin plus rroderator surrounded by a cuboid of water equal in height to the sleeve; elevations are based on z4 at the B ASKET bottom. Fuel pin and moderator dimensions are fuel assembly type dependent.

Unit KENO-Va input Moderator 7

e Cladding Gap ~

l y

~

7 ZCYLINDER FuelMaterial fuel Radius Active FuelIIcight ZCYLINDER Gap Material Gap Radius Fuel Rod-Fuel Rod Erid Cap licight ZCYLINDER Clad Material Clad Radius Fuel Rodlicight CUBOID Moderator "4P"-lialf Fuel Pitch Sleeve licight Unit 18 Water IIole A water hole to be used in those BWR assembies containing inert tubes or water rods, Water p

Unit KENO-Ya input CUBOID Moderator "4P"-IIalf Fuel Pitch Sleeve IIcight e

Revis6on Prepared Date Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 B-2 W

TranStor" BWR Criticality Analysis c ' P 8i

""N Calcu!abon 14 umber lilit L1,10 E M i

SNC Sierra Nuclear Corporation Unit 11 X-Direction Poison Sheet Poison Sheet to be placed between assembly sheets (Sheets required for X and Y coordinate planes).

W ater AlClad pna 4i ene.

[

• f,

.1 l

y_.

'8

+

.4, wi t.

o b&s Unit KENO-Va input CUBOID DORAL 2Pw./2 2Pt./2 Elevations (2 & 3 inch from ends of sleeve)

CUBOID Aluminum 2Pw./2 2Pt /2 Elevations (2 & 3 inch from ends of sleeve)

CUBOID Water 2Pw./2 2Pt/2 Elevations (2 & 3 inch from ends of sleeve)

CUBOID Steel 2P b&s/2 2Pt/2 Elevations (2 & 3 inch from ends of sleeve)

Unit 12 Y-Direction Poison Sheet Y-Direction Poison Sheet. Identical to Unit 5 except for reversal of x and y coordiantes.

Unit 13 Steel Block Corner block to fill array sleeves and spacer /BORAL sheet Unit cells.

Steel I

t, Unit KENO-Ya Input CUBOID Water 4Pt,2 Sleeve Elevations l

Revts40n Prepared ate Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 B-3 Subject TranStor"BWR Criticality Analysis M P-S

""" 8*8

~ Calculation Number difi:Lt.10.OG.1b

SNC Sierra Nuclear Corporation Unit 10 Fuel Assembly Centered in Sleeve Opening i

The fuel pin array surrounded by a water cube, assembly channel and channel to sleeve water gap (if the fuel assembly is modeled with channel) and the assembly sleeve. The Units center is located at the midpoint of the sleeve. Fuel pin and channel lower left coordinates are adjusted to move the fuel assembly within the sleeve. Unit 10 represents the fuel assembly centered in the channel and sleeve. This Unit basic structure is repeated 8 times to model fuel movement towards the top, bottom, ten and right sleeve walls, and the top left and right comer and bottom left and right sleeve corner positions Note: In this revition of the document (as of Rev. 4), a fuel pin array to sleeve gap of zero is assumed for all cases.

Fuel Pin Array to 7

/

ChannelGap

/

Fuel Assembly Channel

{

/

/

Channelto Fuel Pin Array

/

Fuel Assembly l

Sleeve

/

Unit KENO-Va input ARRAYI x

y 0.0 CUBOID Water 4P Channel Inner Imwer Left Coordinate FuelHeight End Fitting CUBOID Steel 4P Channel Outer IAwer Ien Coordinate FuelHeight End Fitting CUBOID Water 4P SleeveInner Lower Len Coordinate Sleeve Elevations CUBOID Steel 4P Sleeve Outer Lower Len Coordinate Sleeve Elevation Unit 2-Unit 9 A repeat of Unit 10. A repeat of the unit will allow independent movement of the fuel assemblies w sleeve. A units are constructed from Array 1 (Region 1) and a water cuboid.

Unit 19 Empty Fuel Sleeve y

A repeat of Unit 10 with no assembly array inserted into the central water square. Only used in 60 element basket configurations.

Revision Fivred Date Chemed Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 B-4 Subject TranStor* BWR Criticality Analysis C' I' N "Ev 8 #2*58 Calculation Number BNFLt.10.06.1E I

t i

SNC Sierra Nuclear Corporation UNIT 17 Top Sleeve Combination A unit cell combining the three fuel assembly sleeves with a horizontal and vertical BORAL sheets and steel spacers.

VerticalBo 1 Sheets

/

/

/

Fuel Sleeve Fue Sleeve

)

Fuel Sleeve l

SteelSpacer

\\

Horizonti BORAL Sheets Dimension for this unit is included in Table 6.3.

UNIT 14,15 and 16 i

I Identical to Unit 17 except for orientation. Units 14 through 16 model the three sleeve arrays to be placed at the left, right and bottom of the center 7x7 array.

UNIT 20 GLOBAL UNIT Basket and Shipping Cask model. See data in Table 3.1.

e-l che<Aed Date Sheet Revtsion Prepared pate Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 B-5 Subject TranStor"BWR Criticality Analysis e ' t' '

LISl-L1.10.OG.1E-

• • • • • • • • • • 8 Calculation idumber

SNC Sierra Nuclear Corporation APPENDIX C BORAL MATERIAL DESCRIPTION J

l f

<+

Revision Prepared Date Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 C-1 Subject TranStor" BWR Criticality Analysis cf C -

~

~

• • • * ' N ' Calculation Number BNFL).10.06.16 I

j

SNC i

Sierra Nuclear Corporation

?.

GBS

-2 EX U J-5

-9 m

p,S H

H H

d d.

J.

l Revision Prepared Date Cheded Date Sheet CNent/ Project BNFL

/

4 JEH 8/4/99 KDW 8/6/99 C.2 W

TranStor"BWR Criucality Analysis of N w e.motam8 ~Calculabon (4 umber B(4FLt.10.00.15 i

SNC

1. ~""

APPENDIX E

)

DISK CONTAINING COMPLETEINPUT AND OUTPUT LISTINGS e

Revisnon Prepared Date Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 E-1 EW TranStor* BWR Criticality Analysis of E-7

<wa u. uv o m.w Calculation Number bNF L1,10.0G,1 L

z SNC Sierra Nuclear Corporation I

l I

t l

i Rev6sion Prepared Date Checked Date Sheet Client / Project BNFL 4

JEH 8/4/99 KDW 8/6/99 E-2 Subject TranStor" BWR Criticality Analysis oW wm.acv a ma ~calcutation taumLet u n u.w.co m

. l ENCLOSURE 9 REPLACEMENT PAGES FOR VOLUME IV, TAB 7 TABLE I COMPLIANCE MATRIX TO NUREG-1617

7 BNFL Fuel Solutions BFS/NRC 99-083. Enclosure 9 LIST OF INSTRUCTIONS REMOVE PAGE(S)

INSERT PAGE(S)

Pages 1 through 7 Pages 1 through 7 1

i

[

TABLE 1 COMPLI ANCE MATRIX TO NUREG-1617 NUREG-1617 Guidance TranStor" Compliance

1. GENERAL INFORMATION REVIEW 1.1 REVIEW OBJECTIVE Yes 1.2 AREAS OF REVIEW Yes 1.2.1 General SAR Format Yes 1.2.2 Package Design Information Yes 1.2.3 Package Description Yes 1.2.4 Compliance with 10 CFR Part 71 Yes 1.2.5 Appendix Yes 1.3 REGULATORY REQUIREMENTS Yes 1.3.1 General SAR Format See Table 2 1.3.2 Package Design Information See Table 2 1.3.3 Package Description Yes 1.3.4 Compliance with 10 CFR Part 71 Yes 1.4 ACCEPTANCE CRITERIA Yes 1.4.1 General SAR Format See Table 2 1.4.2 Package Design Information Yes 1.4.3 Package Description Yes 1.4.4 Compliance with 10 CFR Part 71 Yes 1.5 REVIEW PROCEDURES Yes 1.5.1 General SAR Format See Table 2 1.5.2 Package Design Information See Table 2 1.5.3 Package Description See Table 2 1.5.4 Compliance with 10 CFR Part 71 See Table 2 1.5.5 Appendix Yes 1.6 EVALUATION FINDINGS Yes

1.7 REFERENCES

Yes j

2. STRUCTURAL REVIEW 2.1 REVIEW OBJECTIVE l

Yes

2.295 AREAS OF REVIEWA

- *< v 2.2.1 Description of Structural Design Yes i

2.2.2 Material Properties Yes 2.2.3 Lifting and Tie-down Standards for All Packages Yes 2.2.4 General Considerations for Structural Evaluation of Packaging Yes l

2.2.5 Normal Considerations of Transport Yes 2.2.6 Hypothetical Accident Conditions Yes 2.2.7 Special Requirement for Irradiated Nuclear Fuel Shipments Yes 2.2.8 Internal Pressure Test Yes 2.2.9 Appendix Yes i

- Page 1-

TABLE 1 COMPLI ANCE MATRIX TO NUREG-1617 NUREG-1617 Guidance TranStor Compliance

<2.3 a REGULATORY REQUIREMENTS w

o 2.3.1 Description of Structural Design Yes 2.3.2 Material Properties Yes 2.3.3 Lifting and Tie-down Standards for All Packages Yes 2.3.4 General Considerations for Structural Evaluation of Packaging Yes 2.3.5 Normal Considerations ofTranspon Yes 2.3.6 Hypothetical Accident Conditions Yes 2.3.7 Special Requirement for Irradiated Nuclear Fuel Shipments Yes 2.3.8 Internal Pressure Test Yes 2.4 % : ACCEPTANCE CRITERIA e

^

C -

2.4.1 Description of Structural Design See Table 2 2.4.2 Material Propedies See Table 2 2.4.3 Lifting and Tie-down Standards for All Packages Yes 2.4.4 General Considerations for Structural Evaluation of Packaging Yes 2.4.5 Normal Considerations ofTransport Yes 2.4.6 Hypothetical Accident Conditions Yes 2.4.7 Special Requirement for Irradiated Nuclear Fuel Shipments Yes 2.4.8 Internal Pressure Test Yes 2.54-

? REVIEW PROCEDURES u '

2.5.1 Description of Stmetural Design See Table 2 2.5.2 Material Properties See Table 2 2.5.3 Lining and Tie-down Standards for All Packages Yes 2.5.4 General Considerations for Structural Evaluation of Packaging Yes 2.5.5 Normal Considerations ofTransport See Table 2 2.5.6 Hypothetical Accident Conditions See Table 2 2.5.7 Special Requirement for Irradiated Nuclear Fuel Shipments Yes 2.5.8 Internal Pressure Test Yes 2.5.9 A 3pendix See Table 2 2XiRWRhllWUATIONFINDINGS&Mu%WeVWMa* @ m N$M 2.6.1 Description of Structural Design See Table 2 2.6.2 Material Propedies Yes 2.6.3 Lifting and Tie-down Standards for All Packages Yes 2.6.4 General Considerations for Structural Evaluation of Packaging Yes 2.6.5 Normal Considerations ofTranspon Yes i

2.6.6 Hypothetical Accident Omditions Yes j

._2.6.7 Special Requirement for leadiated Nuclear Fuel Shipments Yes

)

2.6.8 Internal Pressure Test Yes

2.7 REFERENCES

See Table 2

3. THERMAL REVIEW 3.1 REVIEW OBJECTIVE l

Yes

@ S L3.2M ' AREAS OFREVIEWh 3.2.1 Description of the Thennal Design l

Yes

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TABI.E1 COMPl IANCE M ATRIX TO NUREG-1617 NUREG-1617 Guidance TranStor" Compliance 3.2.2 Material Properties and Component Specifications Yes 3.2.3 Thermal Evaluation Methods Yes 3.2.4 Evaluation of Accessible Surface Temperatures Yes 3.2.5 Evaluation under Normal Conditions of Transport Yes 3.2.6 Evaluation under Hypothetical Accident Conditions Yes 3.2.7 Appendix Yes 3.3 :

REGULATORY REQUIREMENTS 3.3.1 Description of the Thermal Design See Table 2 3.3.2 Material Properties and Component Specifications See Table 2 3.3.3 Thermal Evaluation Methods Yes 3.3.4 Evaluation of Accessible Surface Temperatures Yes 3.3.5 Thermal Evaluation under Normal Conditions of Transport Yes 3.3.6 Thermal Evaluation under Hypothetical Accident Conditions Yes

'3.45 WACCEPTANCE CRITERIAX,

N ' re #F, av 3.4.1 Description of the Thermal Design Yes 3.4.2 Material Properties and Component Specifications Yes 3.4.3 Thermal Evaluation Methods See Table 2 3.4.4 Evaluation of Accessible Surface Temperature Yes 3.4.5 Thermal Evaluation under Normal Conditions of Transport Yes 3.4.6 Thennal Evaluation under Hypothetical Accident Conditions Yes

'3.5a 1 REVIEW PROCEDURES '

U 3.5.1 Description of the Thermal Design See Table 2 3.5.2 Material Properties and Component Specifications See Table 2 3.5.3 Thermal Evaluation Methods See Table 2 3.5.4 Evaluation of Accessible Surface Temperatures Yes 3.5.5 Thermal Evaluation under Normal Conditions of Transport See Table 2 3.5.6 Thermal Evaluation under Hypothetical Accident Conditions See Table 2 3.5.7 Appendix See Table 2 1,9!@$VAI11ATIONFINDINGSP vie WMMWWWW WEE $sWW 3.6.1 Description of the Thennal Design Yes j

3.6.2 Material Properties and Component Specifications Yes 3.6.3 Thermal Evaluation Methods Yes 3.6.4 Evaluation of Accessible Surface Temperatures Yes 3.6.5 Evaluation under Normal Conditions of Transport Yes 3.6.6 Evaluation under Hypothetical Accident Conditions Yes

3.7 REFERENCES

Yes j

4. CONTAINMENT 4.1 REVIEW OBJECTIVE Yes 4.2 AREAS OF REVIEW Yes 4.2.1 Description of Containment System See Table 2 4.2.2 Containment under Normal Conditions of Transport Yes 4.2.3 Containment under Hypothetical Accident Conditions Yes

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I TABLE 1 COMPLI ANCE MATRIX TO NUREG-1617 NUREG-1617 Guidance TranStor" Compliance 4.2.4 Appendix Yes 4.3 REGULATORY REQUIREMENTS Yes 4.3.1 Description of Containment System See Table 2 4.3.2 Containment under Normal Conditions of Transport Yes 4.3.3 Containment under Hypothetical Accident Conditions Yes 4.4 ACCEPTANCE CRITERIA Yes 4.4.1 Description of Containment System See Table 2 4.4.2 Containment under Normal Conditions of Transport Yes 4.4.3 Containment under Hypothetical Accident Conditions Yes 4.5 REVIEW PROCEDURES Yes 4.5.1 Description of Containment System See Table 2 4.5.2 Containment under Nonnal Conditions of Transport See Table 2 4.5.3 Containment under Hypothetical Accident Conditions See Table 2 4.5.4 Appendix Yes 4.6 EVALUATION FINDINGS Yes 4.6.1 Description of Containment System See Table 2 4.6.2 Containment under Normal Conditions of Transport Yes 4.6.3 Containment under Hypothetical Accident Conditions Yes

4.7 REFERENCES

Yes

5. SHIELDING REVIEW 5.1 REVIEW OBJECTIVE Yes 5.2 AREAS OF REVIEW Yes 5.2.1 Description of the Shielding Design Yes 5.2.2 Source Specification Yes 5.2.3 Model Specification See Table 2 5.2.4 Evaluation Yes 5.2.5 Appendix Yes 5.3 REGULATORY REQUIREMENTS Yes 5.3.1 Description of the Shielding Design See Table 2 5.3.2 Source Specification Yes 5.3.3 Model Specification Yes 5.3.4 Evaluation Yes 5.4 ACCEPTANCE CRITERIA Yes 5.4.1 Description of the Shielding Design Yes 5.4.2 Source Specification Yes j

5.4.3 Model Specification See Table 2 5.4.4 Evaluation Yes 5.5 REVIEW PROCEDURES Yes 5.5.1 Description of the Shielding Design Yes 5.5.2 Source Specification Yes 5.5.3 Model Specification See Table 2 5.5.4 Evaluation Yes i

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e

- ?

TABLE 1 l

COMPLIANCE MATRIX TO NUREG-1617 NUREG-1617 Guidance TranStor" Compliance l

5.5.5 Appendix Yes 5.6 EVALUATION FINDINGS Yes _

5.6.1 Description of the Shielding Design Yes 5.6.2 Source Specification Yes 5.6.3 Model Specification Yes 5.6.4 Evaluation Yes

5.7 REFERENCES

Yes 6.0 CRITICALITY REVIEW

]

6.1 REVIEW OBJECTIVE Yes 6.2 AREAS OF REVIEW Yes 6.2.1 Description of Criticality Design Yes 1

6.2.2 Spent Nuclear Fuel Contents Yes 6.2.3 General Considerations for Evaluations Yes 6.2.4 Single Package Evaluation Yes 6.2.5 Evaluation of Package Arrays under Nornial Conditions of Yes Transport 6.2.6 Evaluation of Package Arrays under Hypothetical Accident Yes Conditions 6.2.7 Benchmark Evaluations Yes 6.2.8 Appendix Yes 6.3 REGULATORY REQUIREMENTS Yes 1

6.3.1 Description of Criticality Design Yes 6.3.2 Spent Nuclear Fuel Contents Yes 6.3.3 General Considerations for Evaluations Yes 6.3.4 Single Package Evaluation Yes j

6.3.5 Evaluation of Package Arrays under Nonnal Conditions of Yes Transport 6.3.6 Evaluation of Package Arrays under Hypothetical Accident Yes Conditions 6.3.7 Benchmark Evaluations Yes 6.4 ACCEPTANCE CRITERIA Yes 6.4.1 Description of Criticality Design Yes 6.4.2 Spent Nuclear Fuel Contents Yes 6.4.3 General Considerations for Evaluations Yes 6.4.4 Single Package Evaluation Yes 6.4.5 Evaluation of Package Arrays under Normal Conditions of Yes Transport -

6.4.6 Evaluation of Package Arrays under Hypothetical Accident Yes Conditions 6.4.7 Benchmark Evaluations Yes 6.5 REVIEW PROCEDURES Yes 6.5.1 Description of the Criticality Design Yes 4

- Page 5-

)

r m

TABI,E 1 COMPI,I ANCE MATRIX TO NUREG-1617 NUREG-1617 Guidance l TranStor Compliance 6.5.2 Spent Nuclear Fuel Contents Yes 6.5.3 General Considerations for Evaluations Yes 6.5.4 Single Package Evaluation Yes 6.5.5 Evaluation of Package Arrays under Normal Conditions of Yes Transport 6.5.6 Evaluation of Package Arrays under Hypothetical Accident Yes Conditions 6.5.7 Benchmark Evaluations Yes 6.5.8 Appendix Yes 6.6 EVALUATION FINDINGS Yes 6.6.1 Description of Criticality Design Yes 6.6.2 Spent Nuclear Fuel Contents Yes 6.6.3 General Considerations for Evaluations Yes 6.6.4 Single Package Evaluation Yes 6.6.5 Evaluation of Package Arrays under Normal Conditions of Yes Transport 6.6.6 Evaluation of Package Arrays under Hypothetical Accident Yes Conditions 6.6.7 Benchmark Evaluations Yes

6.7 REFERENCES

Yes 7.0 OPERATING PROCEDURES REVIEW 7.1 REVIEW OBJECTIVE l

Yes

7;2N@ AREAS OF REVIEWhbh "L,'

^

?

6 7.2.1 Package Loading Yes 7.2.2 Package Unloading Yes 7.2.3 Preparation of Empty Package for Transport Yes 7.2.4 Other Procedures Yes

!723M$$RE001nTORY4tBQUIREMENTSDd4MMMi@MeWMA WN4 7.3.1 Package Loading Yes 7.3.2 Package Unloading Yes 7.3.3 Preparation of Empty Package for Transport Yes 7.3.4 Other Procedures Yes

~

17;4:

HACCEPTANCE CRITERIAS w

/'

7.4.1 Package Loading Yes 7.4.2 Package Unloading Yes 7.4.3 Preparation of Empty Package for Transport Yes 7.4.4 Other Procedures Yes

!7.5WaREVIEW.PROCEDURESWFM wm emn MMak y., n #n 7.5.1 Package Loading Yes 7.5.2 Package Unloading Yes 7.5.3 Preparation of Empty Package for Transport Yes 1

7.5.4 Other Procedures Yes

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/

i TABLE 1 COMPLI ANCE MATRIX TO NUREG-1611 i

NUREG-1617 Guidance TranStor Compliance 7.6, w EVALUATION FINDINGS"

, Mr'

/

3

+

7.6.1 Package loading Yes 7.6.2 Package Unloading Yes 7.6.3 Preparation of Empty Package for Transport Yes 7.6.4 Other Procedures Yes 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW 8.1 REVIEW OBJECTIVE l

Yes 18.2 r l-

• ACCEPTANCE TESTSw To e

'4

~

J-8.2.1 Areas of Review Yes 8.2.2 Regulatory Requirements Yes 8.2.3 Acceptance Criteria Yes 8.2.4 Review Procedures Yes 8.2.5 Evaluation Findings Yes

-8;3WPMAINTENANCEPROGRAM, mW: h te ud P-S 8.3.1 Areas of Review Yes 8.3.2 Regulatory Requirements Yes 8.3.3 Acceptance Criteria Yes 8.3.4 Review Procedures Yes 8.3.5 Evaluation Findings Yes APPENDIX A-STANDARD REVIEW PLAN CORRELATION Yes WITH 10 CFR PART 71 AND REGULATORY GUIDE 7.9 APPENDIX B -TABLE OF EXTERNAL DOSE RATES FOR See Table 2 EXCLUSIVE-USE SHIPMENTS i

i

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