ML20211H882
| ML20211H882 | |
| Person / Time | |
|---|---|
| Site: | 07109270 |
| Issue date: | 08/30/1999 |
| From: | Mcginty T NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Thompson T NAC INTERNATIONAL INC. (FORMERLY NUCLEAR ASSURANCE |
| References | |
| TAC-L22452, NUDOCS 9909020201 | |
| Download: ML20211H882 (40) | |
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[ 4 UNITED STATES u E NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 4 001 4
9***** August 30,1999 Mr. Thomas C. Thompson, Director Licensing & Competitive Assessment NAC International, Inc.
655 Engineering Drive Norcross, GA 30092 '
SUBJECT:
REQUEST FOF ADDITIONAL INFORMATION FOR THE UMS UNIVERSAL TF ANSPORT SYSTEM (TAC NO. L22452)
Dear Mr. Thompson:
]
By application dated April 30,199), as supplemented on December 24,1997, April 23,1998, and June 18,1999, NAC Internationd inc. (NAC) requested approval of the UMS Universal ;
Transport System under the provisions of 10 CFR Part 71. Enclosed is the staff's request for i additionalinformation (RAl) for the continued review of the UMS Universal Transport System.
Your response to the enclosed RAI is expected by December 1,1999, as identified in the current review schedule. If you are unable to meet the established RAI response milestone, you must notify us in writing, at least 2 weeks prior to the due date, of your new submittal date and the reasons for the delay. We will then assess the impact of the new submittal date and publish a revised schedule.
If you have any comments or questions concerning this request, you may contact me at <
(301) 415-8580. Please refer to Docket No. 71-9270 and TAC No. L22452 in future correspondence related to this request.
Sincerely, w.cN Y 7W V Timothy J. McGinty, Project Manager f Spent Fuel Licensing Section Spent Fuel Project Office j Office of Nuclear Material Safety I and Safeguards Docket No.: 71-9270 0
Enclosure:
RAI on NAC-UMS Transport cc: Mr. Len Tremblay Mr. Scott Bauer Yankee Atomic Electric Company Arizona Public Service Company Mr. George Zinke Mr. Paul Bemis Maine Yankee Atomic Power Company Stone and Webster Eng. & Construction d
9909020201 990030 PDR ADOCK 0710 2 0 C
J
a e August 30,1999 Mr. Thomas C. Thompson, Director Licensing & Competitive Assessment NAC International, Inc.
655 Engineering Drive Norcross, GA 30092
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE UMS UNIVERSAL TRANSPORT SYSTEM (TAC NO. L22452)
Dear Mr. Thompson:
By application dated April 30,1997, as supplemented on December 24,1997, April 23,1998, and June 18,1999, NAC International, Inc. (NAC) requested approval of the UMS Universal Transport System under the provisions of 10 CFR Part 71. Enclosed is the staff's request for additional information (RAI) for the continued review of the UMS Universal Transport System.
Your response to the enclosed RAI is expected by December 1,1999, as identified in the current review schedule. If you are unable to meet the established RAI response milestone, you must notify us in writing, at least 2 weeks prior to the due date, of your new submittal date and the reasons for the delay. We will then assess the impact of the new submittal date and publish a revised schedule.
If you have any comments or questions concerning this request, you may contact me at (301) 415-8580. Please refer to Docket No. 71-9270 and TAC No. L22452 in future .
correspondence related to this request. l I
Sincerely, Timothy J. McGinty, Project Manager Spent Fuel Licensing Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No.: 71-9270
Enclosure:
RAI on NAC-UMS Transport cc: Mr. Len Tremblay Mr. Scott Bauer Yankee Atomic Electric Company Arizona Public Service Company Mr. George Zinke Mr. Paul Bernis Maine Yankee Atomic Power Company Stone and Webster Eng. & Construction Qstribution:
Docket NRC File Center PUBLIC e NMSS r/f SFPO r/f RParkhill WHodges EWBrach ELeeds SO'Connor VLTharpe SFShankrnan
'See previous concurrence OFC SFPO C SFPO SFPO SFPO SFPO NAME TJMcGinty* EZiegler* DCarlson* KGruss' EKeegan*
DAM 8 / 27 s 9 8 / 27 S 9 8 / 27 S9 8 / 27 S 9 8 / 27 M9 OFC SFPO SFPO SFPO SFPO, SFPO NAME SColpo* DTang* EEaston* CR e DATE 8 / 27/99 8 / 27/99 8 / 27/99 / O /99 / /99 C = Cover E = Cover & Enclosure N = No copy OFFICIAL RECORD COPY G:\NAC\UMST\RAl 1\RAl1CVR
1 s
Mr. T. C. Thompson 2 l The outlined conditions for a CoC should include: 3 package description; a list of licensing drawings with appropriate revisions; tables and/or descriptions which define the type, form, and maximum quantities of the allowable contents, including variable cooling times and GTCC waste; descriptions of the appropriate provisions from the operating procedures, aintenance procedures, fabrication controls, and acceptance tests; and any other appropri e conditions necessary to ensure compliance with 10 CFR Part 71.
If you have any comments or questions concerning this request, you may ontact me at (301) 415-8580. Please refer to Docket No. 71-9270 and TAC No. L22 2 in future ;
correspondence related to this request.
Sincerely, Timothy J. McGinty, Project Manager Spent Fuel Li nsing Section Spent Fuel P oject Office Office of Nu lear Material Safety a d Safeguards Docket No.: 71-9270
Enclosure:
RAI on NAC-UMS Transp rt cc: Mr. Len Tremblay Mr. Scott Bauer Yankee Atomic Electric Comp ny Arizona Public Service Company Mr. George Zinke Mr. Michael Meisner Maine Yankee Atomic Pow r Company Maine Yankee Atomic Power Company Mr. Paul Bemis Stone and Webster Eng. Construction Distribution:
Docket NRC F e Center PUBLIC NMSS r/f SFPO r/f RParkhill WHodges EWBr ch ELeeds SO'Connor VLTharpe SFShankman OFC SFPO h SF m b SFPO h. / b SFPO b SFPO NAME DCarlson j KGruss M k- EKeegan DATE [ M/99 h 99 / /99 b /13 /99 99 OFC SFPO SFPO / M , b SFPO ,, fyh SFPO SFPO NAME SColpo \e# DTang ( AA /_ m EEaston CRChappell DATE / /99 / /99 b/ 99 / /99 / /99 C a Cover E e Cover & Enclosure N = No copy OFFICIAL RECORD COPY G.WAC\UMSTWAl 1\RAllCVR
g
- .L s NAC UMS UNIVERSAL TRANSPORT CASK DOCKET NO. 71-9270 l TAC NO. L22452 REQUEST FOR ADDITIONAL INFORMATION 1 This document titled Request for Additional Information (RAl), contains a compilation of additional information requirements, Identified to-date by the U.S. Nuclear Regulatory Commission (NRC) staff, during its first round review of NAC International's application for approval of the NAC Universal Multi-Purpose Canister (NAC-UMS) transport system under 10 CFR Part 71. This RAI follows the same format as NAC's Safety Analysis Report (SAR).
Each individual RAI describes information needed by the staff for it to complete its review of the l application and the SAR and to determine whether NAC has demonstrated compliance with the
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regulatory requirements.
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ENCLOSURE Page 1 of 37
CHAPTER 1 GENERAL INFORMATION Section 1.0 ' Generalinformation 1-1 - Remove damaged fuel from the site-specific fuel terminology in Table 1-1, or provide l the appropriate analysis to support including damaged fuel.
. The site-specific fuel terminology from Table 1-1 includes containerized damaged fuel.
Analysis has not been presented for damaged fuel and, therefore, cannot be included as site-specific fuel.
Section 71.7(a) requires complete and accurate information. The inclusion of damaged fuel in this Table implies that analysis for damaged fuel has been included in the SAR.
1-2 Specify Zircaloy clad fuel rods in the Standard Fuel description.
Zircaloy clad rods were not specified in the Standard Fuel description. The loading tables are based on Zircaloy clad fuel. Section 71.7(a) requires complete and accurate information.
13 Remove the statement on pg.1-1 of the application to clarify that the transport index for nuclear criticality was based upon both the optimum internal moderation and the i optimum external moderation.
In justifying a transport index of zero, NAC's analysis has shown that an infinite number l of packages with optimum external and internal moderation would remain suberitical.
Section 71.59 establishes the requirements for determining the transport index based on the evaluation of package arrays for normal and accident conditions. Standard Review 3
Plan (SRP) Section 6.5.6.1 states that the evaluation of accident arrays should consider '
the most reactive configurations of water inleakage and internal moderation.
Section 1.2.1.2.1 Cask Body 1-4 Describe the function and configuration of the heat transfer fins in the neutron shield.
There is no description of the fins or heat transfer function in this Section. The description should include dimensions of the fins and material composition. Section 71.43(f) provides that a package must be designed to limit radiation exposure and ensure there is no substantial reduction in the effectiveness of the package. Section 71.33 and SRP Section 1.5.3.2 describe that the application must include a description which contains a sufficient basis for evaluation.
Section 1.2.1.2.9 Transportable Storage Canister Cask Cavity Spacers 1-5 Verify the lengths of the Universal Transport Cask spacers for each of the five fuel class
- canister / spacer configurations. Also, verify that the spacer (s) and Transportable Storage Canister (TSC), or the Greater Than Class C (GTCC) waste canister, will fit into the cavity of the Universal Transport Cask. Change Drawing No. 790-520 and/or the SAR text, as appropriate, to reflect the correct spacer length.
Page 2 of 37
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Section 71.7(a) requires that the SAR contain complete and accurate information. This
. SAR Section indicates that the lengths of the Class 1 and Class 2 pressurized water reactor (PWR) spacers are 7.67 and 16.75 inches, respectively. However, there appears to be an inconsistency with Drawing No. 790-520, which shows the lengths to be 11.25 and 18.25 inches.
Section 1.2.3 Contents of Packaging 1-6 Clarify whether unenriched fuel assemblies are intended to be acceptable contents.
The first sentence of the third paragraph from the bottom of page 1.2-14 is ambiguous with respect to the treatment of unenriched fuel assemblies. Section 71.7(a) requires complete and accurate information.
1-7 include the minimum initial enrichment in the table (pg.1.2-15) that describes cask spent fuel contents. Additionally, Tables 1.2-4 and 1.2-5 list only maximum enrichments and also should include minimum enrichments.
The minimum enrichment, in combination with other specifications, is a parameter on which the bounding source term and heat load are determined. As such, it should be included in the SAR. Section 71.33 requires that the application include a description with sufficient detail to identify the package accurately and provide a sufficient basis for l
, evaluation of the package.
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1-8 include a footnote for clarification and consistency between the table on pg.1.2-15 which states that the minimum cooling time is 6 years and Table 1.2-6, which lists a minimum cool time of 5 years for high enriched, low burnup fuel.
I Section 71.7(a) requires complete and accurate information. 1 1-9 Justify the shipment of fuels with a burnup greater than 45,000 mwd /MTU. The justification should include the quantitative effects zirconium hydriding may have on the mechanical properties of the cladding (i.e., tensile strength, yield strength, ductility, fracture toughness, etc.) under normal conditions of transport. Consideration should be given to the loads on the cladding associated with both the vibration normally incident to transport and the normal condition free drop from 1 foot. Consideration should also be given to (1) the potential for dissolution of any existing zirconium hydrides during the short-term higher temperatures encountered during transportation, (2) the potential for ,
re-precipitation and/or re orientation of the hydrides if the temperature of the cladding 1 decreases during transportation, and (3) the impact that zirconium hydrides may have on cladding integrity under normal transport condition loads.
Section 71.55(d)(2) requires that the geometric form of the package contents of a spent fuel package will not be substantially altered under the conditions specified,for normal conditions of transport. The staff needs this information to complete the containment review required by Section 71.51. The amount of hydrogen picked up by the Zircaloy cladding during reactor operation may effect the mechanical properties of high burnup fuel prior to and during transportation. Under normal conditions of transport, the loads associated with vibration and/or a free drop from 1 foot may impart loads that could Page 3 of 37
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disrupt the integrity of the fuel cladding and lead to an unanalyzed containment I situation.
1-10 Explain the reason for the two different minimum cooling times. In this Section, the minimum cooling time is stated as 6 years, however, in the shielding analyses Chapter the analyses for design basis fuel uses 10 years minimum cooling time.
The different cooling times combined with various fuel burnups will result in different source terms. Section 71.33 requires the identification of the maximum radioactivity of the package constituents.
1-11 Provide a table, suitable for inclusion in a Certificate of Compliance, that tabulates the authorized contents of the package. The table should address all requested contents, including variable minimum cooling times, maximum decay heat per assembly, Maine Yankee site-specific fuels (see related RAI 1-12), and the type, form, and maximum quantities of GTCC waste. The contents specified in the table should be consistent with the thermal, shielding, and criticality analyses performed for the package.
Section 71.33(b) requires the application to include a description of the proposed contents in sufficient detail to provide a sufficient basis for evaluation of the package.
SRP Section 1.5.2.3 provides guidance that the contents in the package description are sufficient to allow an understanding of exactly what is to be transported.
1-12 Provide a table, suitable for inclusion in a Certificate of Compliance, that tabulates the authorized contents of the Maine Yankee site-specific contents. The table may reference the principal characteristics of the standard design basis 14 x 14 fuel assemblies. The table should establish the differences from the standard design basis assemblies, include burnup specifications and the applicability of the variable cooling time loading tables, and define the configurations of the site-specific fuel assemblies that have been shown by analysis to be acceptable.
Section 71.33(b) requires the application to include a description of the proposed contents in sufficient detail to provide a sufficient basis for evaluation of the package.
SRP Section 1.5.2.3 provides guidance that the contents in the package description are sufficient to allow an understanding of exactly what is to be transported.
Tables 1.2-6 & -7 Loading Tables for PWR and BWR Fuels 1-13 Provide additional information on the validity of these tables. These tables identify fuels with a wide range of burnup and cool times that could be loaded into the cask.
Additionally, a cool time of 5 years is allowed.
For example, this time frame is inconsistent with information provided in Section 1.2.3 and the design basis fuel identified in Chapter 5. Section 71.43(f) provides that a package must be designed to limit radiation exposure and ensure there is no substantial reduction in the effectiveness of the package.
Page 4 of 37
- Section 1.3.1.1.1 ' Maine Yankee Site Specific Spent Fuel Configurations 1-14 Provide a description of how these spent fuel assemblies have been determined to be acceptable with respect to maximum activity.
Page 1.3.1-3 of the SAR lists site-specific fuel assembly configurations that meet the criteria for acceptable contents. However, no analysis is presented or referenced.
- Section 71.33 requires the identification of the maximum radioactivity of the package constituents.
Section 1.3.4 - License Drawings 1-15 Include detail E-E from Drawing No. 790-502.
If detail E-E does not show the Cu-SS fins, include an annular cross-sectional view of the cask body, including the fins and their relationship with surrounding components, including the lead and neutron shields. Section 71.33 requires sufficient package detail to provide a basis for evaluation of the package.
1-16 Clarify the term "Out-of-Spec" with respect to boiling water reactor (BWR) fuel tubes.
Update the appropriate Sections of the SAR to explain the meaning.
Section 71.7(a) requires that the SAR contain complete and accurate information. The term "Out-of Spec"is used on D. awing Nos. 790-501 and 790-605, but the meaning of the term is not explained.-
1-17 Clarify which paint coating will be applied to the BWR support disks. Revise the appropriate SAR Sections, the Drawings, or both, to reflect the correct choice of paint coating.
Section 71.7(a) requires that the SAR contain complete and accurate information. '
Drawing No. 790-573 indicates that either a Keeler & Long Hi-Heat Silicone Aluminum No. 3731 or an Ameron Amercoat 878 silicone coating will be applied. However, this ,
specification is inconsistent with the text of Section 2.4.4.2.10 which indicates that an Amercoat PSX 738 coating will be applied.
1-18 Revise Drawing No. 790-502, sheet 4, by adding a note on the torque value for item 26 (threaded inserts) for bolting the secondary lifting trunnions to the cask.
Complete and accurate information should be listed on licensing drawings, per Section 7.1.7(a).
1-19 Revise Drawing No. 790 502, sheets 3 and 4, by noting the filler material grades used for welding the shear rings, trunnions, and rotation pockets to the cask.
.The structural analyses of SAR Section 2.5, which demonstrates compliance with 71.45(b)(3), relies on the consideration of specific weld strengths to ensure that, under
- excessive load, failure of tie-down devices would not impair the ability of the package to
- meet other 10 CFR Part 71 requirements.
Page 5 of 37 l
1-20 Verify thicknesses for the BORAL panels used in the BWR and PWR baskets, as shown
. in Drawing Nos. 790-575 and 790-581.
NAC's response should indicate the correct panel thickness, the effective thickness of the poison meat, and the volume fraction, boron enrichment, and size distribution of the B,C particles. A BORAL panel of the thickness (0.075 inch) and minimum poison loading (0.025 g ' B/cm 2) specified for the PWR baskets would appear to call for an impossibly large volume fraction of B,C particles in the poison meat. This, and the fact that the BWR-basket panels are thicker (0.135 inch) with a lower poison loading (0.011 g ' B/cm 2), suggest that the panels in the PWR basket may need to be thicker than shown in Drawing No. 790-581.
Section 71.33(a)(5)(ii) requires an accurate description of neutron absorbers. SRP Section 6.5.1.1 calls for reviewing the dimensions and concentrations of neutron poisons.
1-21 Revise Drawing No. 790-509 to reflect a package identification number with a "-85" suffix.
Section 71.85(c) requires the durable marking, prior to first use, of the package with the assigned package identification number. SRP Section 1.5.2.4 specifies that new transportation packages for spent fuel will be assigned a "-85" suffix.
1-22 The application contains 32 fabrication drawings. The drawings show many details and dimensions which are not important to safety. The application should be revised to include a only those engineering drawings that are suitable for reference in a Certificate of Compliance. The drawings should show the general arrangement of the package,its principal features, and the design of safety related components. The drawings should contain the following information:
- 1. dimensions and tolerances important to safety,
- 2. materials of construction,
- 3. applicable codes and standard for fabrication of safety related components,
- 4. size, type, and location of safety related welds and methods of non-destructive examination,
- 5. torque requirements for closure bolts and other threaded devices,
- 6. total weight of the package, and
- 7. location of the tamper indicating features.
Section 71.33 requires sufficient package detail to provide a basis for evaluation of the package. NUREG/CR-5502," Engineering Drawings for 10 CFR Part 71 Package Approvals," provides additional information for the preparation of drawings.
Page 6 of 37 1
CHAPTER 2 STRUCTURAL
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Section 2.1 Structural Design I i
2-1 Provide a summary description of the codes and standards and their exceptions applicable to the structural design of the GTCC waste canister and basket.
Complete and accurate information on the GTCC waste basket should also be presented, per Section 71.7(a). I Section 2.1.2 Design Criteria 2-2 Update Table 2.1.2-1 to reflect the pertinent ASME Code Exceptions as noted in Table 4-1 of the NAC-UMS Storage SAR for the TSC.
Section 71.7(a) requires complete and accurate information. Drawing Nos. 790-585 and
-612 specify a progressive liquid penetrant nondestructive examination, or an ultrasonic examination, of the TSC and GTCC waste canister structural lid to-shell welds.
However, the Drawings are inconsistent with Table 2.1.2-1which specifies a root and final liquid penetrant examination.
I Section 2.1.2.3 Load Combinations ) 1 2-3 Deiine for the cask cavity the maximum normal operating pressure (MNOP) considered in the following SAR text:
- 1. Pgs. 2.6-4 & 2.6-14, "[T]he loading conditions are: (1) 50 psig internal pressure..."
- 2. Table 2.6.1.1-2,"[N]ormal Design Pressure--Cask.. 25 psig..." j
- 3. Pg. 2.10.2-6, "[A] pressure of 150 psig is used to conservatively envelope the )
normal design pressure of 25 psig for allimpact loadings..." '
- 4. Pg. 8.1-4, "[T]he transport cask containment is hydrostatically tested to 85 psig...The containment maximum normal operating pressure (MNOP) is calculated to be 8.5 psig." j Per Section 71.33(b)(5), the MNOP shall be identified.
I 2-4 Clarify, on Pg. 2.1-9, how for the lead-pour fabrication, -40 F cold test and -20 F ambient thermal stresses are considered in the load combination structural evaluations -
of the cask inner shell.
Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package performance under the conditions and tests of Sections 71.71 and 71.73.
2-5 Clarify, on Pg. 2.1-11, the statement, "[t]he visco-elastic behavior of the lead is considered...in the analysis of cask shell components."
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Page 7 of 37
.. m The Section 2.7.1.5 discussion on lead slump does not appear to refer to the
. assumption of visco-elastic material behavior of the lead for the cask shell stress analysis. Complete and accurate information should be provided, per Section 71.7(a).
2-6 Clarify,in Table 2.1.2-1, the code exception to ASME Article NB 6000, Testing,
"[T]ransportable storage canister cannot have a hydrostatic or pneumatic test performed..."
The statement on pg. 8.1-7, "(T]he canister is conservatively pressure tested..1.2 times the 15 psig design pressure...'" appears to be in disagreement with the exception taken in SAR Table 2.1.2-1. Complete and accurate information should be provided in the SAR, per Section 71.7(a), for evaluating packaging codes and standards compliance, per Section 71.31(c).
Section 2.1.2.5.2 Fatigue 2-7 Revise this Section and Sections 2.6.7.6 and 2.7.1.7.2, as appropriate to also consider bolt pre-loads on the cask lid closure bolts for meeting the fatigue evaluation exemption criteria of ASME NB 3222.4(d) and 3232.3(b), including provisions for fatigue strength reduction factors of NB-3232.3(c) for threaded members.
The closure bolt fatigue life needs to be evaluated, and a lower fatigue strength should be considered for the bolt. The effect of repeated use of the closure bolts under normal conditions of transport, per Section 71.71(a), should be included in the evaluation.
Section 2.2 Weights and Centers of Gravity 2-8 Add a table of calculated weights and centers of gravity of the Universal Transport Cask for transporting the GTCC waste.-
Similar to those for the five classes of design basis fuel, complete and accurate weight and center of gravity information on the GTCC waste basket should also be provided, per Section 71.7(a). They are essential for evaluating the applicability.of the bounding decelerations determined for the packaging under the free drop tests of Section 71.71(c)(7) and Section 71.73(c)(1).
Section 2.3.1 Summary of Materials -
2-9 Correct the gamma shield material to be chemical copper-grade lead instead of chemical lead. Section 71.7(a) requires complete and accurate information.
i l 2-10 Revise the table on pg. 2.3-2 to specify the correct grade of lead that will be used for L gamma shielding.
Section 71.7(a) requires complete and accurate information. This Section does not
, indicate which grade of lead will be used for gamma shielding. In Table 2.3.7-1 and on Drawing No. 790 502, it appears that a " Chemical-Copper Grade" lead will be used.
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y Section 2.3.7 Shielding Materials 2-11 In Table 2.3.7-1, correct the values of the " Tensile Yield Strength". Also, indicate the !
source of the data in the table.
Section 71.7(a) requires complete and accurate information. It appears that the decimal point was inadvertently misplaced.
Section 2.4.4 Chemical, Galvanic, or Other Reactions 12 Demonstrate that the placement of the Universal Transport Cask into the spent fuel pool, when the cask is wet loaded, will not impact safety during fuel loading operations.
Consider the effects of chemical, galvanic, or other reactions between the cask materials, including any coatings, and the spent fuel pool water. The potential for generation of combustible gases should be addressed in this evaluation. Revise the operating procedures to include appropriate controls for detecting the presence, and
- preventing the ignition, of any combustible gases that may be generated during cask loading operations (See RAI 7-5).
Section 71.43(d) requires that a package be made of materials that assure there will be no significant chemical or galvanic reactions. Reaction of the Universal Transport Cask components or coatings with spent fuel pool water may produce hydrogen or other flammable gases. Since the shield lids of the TSC and GTCC waste containers are welded to their shells during fuelloading operations, there is a source of heat that could lead to their Ignition if sufficient flammable gas is present.
13 Evaluate the potential for the generation of combustible or reactive gases in both the TSC containing Maine Yankee site specific contents and the GTCC waste canister. As part of the evaluation, identify any nonmetallic materials that will be contained in either of the canisters. If particulate material will be placed inside either canister, identify the material of the particles and determine the range of particle sizes.
Section 71.43(d) requires that a package be made of materials that assure there will be no significant chemical or galvanic reactions. Reaction of unusual materials (e.g., non-stainless steel or non-Zircaloy materials, plastics, resins, etc.) with spent fuel pool water or the radiation and moderately high temperatures of the cask environment may produce flammable or combustible gases. Since the shield lids of the TSC and GTCC waste canister are welded to their shells during loading operations, there is a source of ,
heat that could lead to ignition if sufficient amounts of flammable or combustible gas are i Section 2.4.4.1 Component Operating Environment ;
2-14 Revise this Section to include wet loading of the Universal Transport Cask, as is indicated on pg. 3.4-44, Section 3.4.7, and pg. 7.5-1, Section 7.5. ,
Sections from Chapters 3 and 7 clearly indicate that the Universal Transport Cask or the I transfer cask is immersed in the spent fuel pool. Section 2.4.4.1 Indicates that the j Universal Transport Cask is dry loaded and is not immersed in the pool. Section 1.2.2 indicates that canister loading is accomplished only by use of the transfer cask. Section 71.7(a) requires complete and accurate information.
Page 9 of 37
Section 2.4.4.2.1 Stainless / Nickel Alloy Steels 2-15 in the appropriate Sections, describe how aluminum bronze ferrules and ethylene glycol are used in the Universal Transport Cask, or remove the references to these components if they are not used.
Section 71.7(a) requires complete and accurate information. References are made to aluminum bronze ferrules and ethylene glycol without a description of how these materials are used in the Universal Transport Cask.
Section 2.4.4.2.3 Nonferrous Materials 2-16 Show that the reaction between the aluminum heat transfer disks of the TSC and the spent fuel pool water is not significant with respect to its impact on safety during fuel loading operations. Revise the operating procedures to include appropriate controls for detecting the presence, and preventing the ignition, of combustible gases during cask loading operations (See RAI 7-5). The potential for generation of combustible gases should be addressed in this evaluation. The evaluation should consider (1) the temperature of the water in the TSC will change during loading and (2) the effects of both irradiation and the contact between the aluminum heat transfer disks and the stainless steel washers,' which are used to position the aluminum disks. The evaluation and conclusions should be supported by calculations, experiment, or applicable data gathered from a literature survey.
Section 71.43(d) requires that a package be made of materials that assure there will be j no significant chemical or galvanic reactions. Reaction of the aluminum heat transfer disks with spent fuel pool water and/or steel components may produce hydrogen in concentrations close to the lower explosive limit of hydrogen. Since the shield lid of the TSC is welded to the shell during fuel loading operations, there is a source of heat that could lead to ignition if sufficient amounts of flammable gas are present.
Section 2.4.4.2.10 Coatings 2-17 Demonstrate that the coating applied to the BWR support disks of the TSC will not
- impact safety during fuelloading operations. Revise the operating procedures to
- include appropriate controls for detecting the presence, and preventing the ignition, of combustible gases during cask loading operations (See RAI 7-5). The potential for generation of combustible gases should be addressed in this evaluation. Also, demonstrate that the coating is not reactive and is adherent when it is exposed to PWR and BWR spent fuel pool water, radiation, and temperatures that are expected during fuel loading operations. Describe the process that was used to select the coating.
Include a brief discussion of the tests and/or analyses that were conducted to qualify
~ these coatings for use in the radiation and moderately high temperature environment.
! Indicate the expected impact of flaking or chipping paint on the structural integrity of the l
BWR support disks. Update all SAR Sections, as appropriate, to include these descriptions.
Section 71.43(d) requires that a package be made of materials that assure there will be
- no significant chemical or galvanic reactions. A potential reaction of the paint coating with spent fuel pool water and/or steel components may produce hydrogen or other Page 10 of 37 L -
flammable gases, or it may cause difficulty with loading the spent fuel into the cask.
. Since the shield lid of the TSC is welded to the shell during fuel loading, there is a source of heat that could lead to ignition if sufficient amounts of gas are present.
Section 2.5 Lifting and Tle-Down Standards 2-18 Justify the use of ultimate shear strength for evaluating the structural performance of the lifting and tie down devices of the package under excessive load.
For consistency, the von Mises failure criterion, which is used in the lifting trunnion and rotation pocket stress analyses, per Section 71.45, should also be considered in the evaluation for excessive load.
Section 2.5.1.1.2 Secondary Lifting Trunnions 2 19 Provide a bolt pre load analysis to be consistent with the bolt stress distribution >
assumption depicted in Figure 2.5.1.1-2.
Complete and relevant information on bolt pre-load should be provided to substantiate the stress distribution assumption for the trunnion attachment bolts. The validity of the structural integrity evaluation of the bolted secondary trunnions, as a lifting device per Section 71.45(a), depends on an adequate amount of pre-load in the bolts.
Section 2.6.1 Heat 2-20 Revise the pressures listed in this Section to be consistent with the thermal Section.
The pressures stated on pgs. 2.6-1 and 2.6-2 are inconsistent with Table 3.4-4.
Section 71.7(a) requires complete and accurate information. !
Section 2.6.1.3 Stress Calculations and Comparison to Allowable Stresses 2-21 Clarify, as appropriate, the underlined term below:
Pgs 2.6-4 and 2.6-14 of the SAR,"[T]he stresses throughout the cask body are calculated for... loading conditions for directiv loaded fuel."
Complete and accurate information should be provided, per Section 71.7(a). .
Section 2.6.2 Cold 2 22 Clarify, as appropriate, the meaning of the underlined term below: j Pg. 2.6-15 of the SAR, "[T]hermal hgj refers to 100 F solar insolation..."
Complete and accurate information should be provided, per Section 71.7(a).
2-23 Re-evaluate stresses in the cask body, as appropriate, by considering the minimum internal pressure in combination with the minimum decay heat load and the minimum ambient temperature.
Page 11 of 37
n The SAR text and Table 2.6.2.1-1 suggest that the pressure and thermal loading
. conditions considered deviate from those typically associated with the cold condition test of the normal conditions of transport, per Section 71.71(b).
Section 2.6.5 Vibration l
2-24 Submit an evaluation of the fatigue strength of the tie-down system of the package.
The vibration induced alternating stresses in the tie-down system should be evaluated under the vibration condition of the normal conditions of transport, per Section 71.71(c)(5).
Section 2.6.7 Free Drop (1 Foot) Cask Body Analysis 2 25 Clarify the text by defining explicitly the free-drop deceleration g-loads used in the bounding analyses of the cask body.
)
Complete and accurate information should be provided, per Section71.7(a) 2-26 in Table 2.6.7.2-1, revise, as appropriate, the stress allowable of 20.0 ksi to be consistent with that of 19.1 ksi of Table 2.6.1.3-1 for shell section 13, and re-evaluate stress margins accordingly.
Shell section 13 is shown to have the lowest stress margin for the cask. The stress allowable in Tables 2.6.7.2-1 and 2.6.1.3-1 are expected to be identical for the shell section under the same ambient temperature of 100 F. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package performance under the Section 71.71 normal conditions of transport.
2-27 With respect to Table 2.6.7.12, justify the use of a higher stress allowable of 29.6 ksi for shell section 22 than that of 19.7 ksi for Section 21 for the same shell material.
Table 2.10.2.2-1 lists an identical temperature of 322.6 F for the two shell sections for stress evaluation. Generally, the same stress allowable should be applicable to the two shell sections in close proximity. Complete and accurate information should be -
provided, per Section 71.7(a), for evaluating the package performance under the Section 71.71 normal conditions of transport.
Section 2.6.7.5 Impact Limiters 2 28 Justify the use of a factor of 0.9, on pg. 2.6-64, for the redwood crush stress-strain curve to account for the suggested negative fabrication tolerance for the impact limiters.
A negative fabrication tolerance, as suggested, may not exist. The use of the factor of 0.9 should only be considered for calculating a bounding cask deceleration by the RBCUBED program. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free-drop performance under Sections 71.71(c)(7) and 71.73 (c)(1).
Page 12 of 37 !
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2-29 Justify the use of a factor of 0.9, on pg. 2.6-65, for the balsa wood crush stress-strain
. curve, at 152 F, to account for the suggested fabrication tolerance for the impact limiters.
)
Reference 37 of the SAR,"NAC-STC Safety Analysis Report," considered the same factor of 0.9, but for a higher temperature of 230 F. Complete and accurate information i should be provided, per Section 71.7(a), for evaluating the package free-drop l performance under Sections 71.71(c)(7) and 71.73(c)(1).
2-30 Clarify, as appropriate, the underlined typographical or. editorial errors:
- 1. Pg. 2.6-72, "[T]able 2.6.7.5-4 shows that at impact angles of 99' and 75", ...the !
secondary impact..."
A comparison of the Emax and E1 values in the table suggests that the SAR _.
statement does not appear to be applicable to the case with an impact angle of 60*. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free-drop performance under Section {
71.73(c)(1). l
- 2. Table 2.6.7.5-4, "[E]nergy... absorbed in second limiter..12*/...."
l The percentage value listed does not appear to be consistent with the other data summarized in the table. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free-drop performance under Section 71.73(c)(1).
I 2-31 Justify, with test or analytical results, that the free drop at a 75 oblique drop angle would give rise to the bounding deceleration of the trailing impact limiter in a slap-down event.
The SAR should provide the basis for the assumption that a 75' oblique drop would produce the largest deceleration, thus, the most limiting and damaging condition, to the package undergoing a secondary impact. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free-drop performance under Section 71.73(c)(1).
~
2 32 Considering the at-temperature impact limiter force-deflection curves in lieu of those i' associated with Figures 2.6.7.5-11 through 2.6.7.5-17 for bounding temperatures, demonstrate that the impact limiter drop test results can adequately be predicted with those calculated by program RBCUBED.
The impact limiter force-deflection curves displayed in the figures apply to the bounding temperatures. The force-deflection curves for the temperature at which the tests were conducted should be considered for evaluating correlation between the test and calculated results. Complete and accurate information should be provided in the SAR, per Section 71.7(a), for evaluating the package free-drop performance under Sections 71.71(c)(7) and 71.73(c)(1).
Page 13 of 37
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2 For the 1-foot free drop results summarized in Table 2.6.7.5-1, explain why the
. calculated deceleration force for the lower impact limiter is larger than that for the upper impact limiter, i For the same impact limiter deformation of 1 inch, the upper impact limiter with a larger backed area than the lower impact limiter should give rise to a higher deceleration force in an end drop event. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free-drop pedormance under Section ,
71.71(c)(7).
2-34 For the corner impact force-deformation curves shown in Figures 2.5.7.5-12 and 2.5.7.5-15 for the lower and upper impact limiters, respectively, explain why the curves ;
are markedly different in shape for initial deformations of about 4 inches or less. I For the identical redwood and balsa wood material properties, the RBCUBED calculated i force-deformation curves are expected to have a similar shape for the essentially 1 identical design for the upper and lower impact limiters. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package free- !
drop performance under Sections 71.71(c)(7) and 71.73(c)(1). -
j l
Section 2.6.12.12 Canister Buckling Evaluation for 1-Foot End Drop 2-35 Describe how the ANSYS dynamic shell analysis was performed for the maximum stresses used in the buckling evaluation of the TSC.
I Complete and accurate information should be submitted for review, per Section 71.7(a).
Sections 2.6.13 & 2.6.15 PWR and BWR Basket Analysis-Normal Conditions of Transport 2-36 With respect to Tables 2.6.12.8-1 and 2.6.14.8-1, for the PWR and BWR canisters, respectively, explain why the minimum stress margins and critical cross sections are shown markedly different from each other (0.02 at Section 2 vs. 0.52 at Section 9) for the top corner-drop.
The PWR and BWR canisters are essentially identicalin design configurations and ;
loading conditions except that the BWR canister is slightly longer. As such, because of !
the same stress analysis approach, minimum stress margins of similar order of magnitude are expected for the same canister cross section locations. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package performance under the Section 71.71 normal conditions of transport.
2-37 Submit the support disk modal properties data to demonstrate that dynamic load factors (DLFs) have appropriately been considered in analyzing support disk ligaments.
The cask deceleration may need to be amplified by dynamic effects for defining the deceleration forces for quasi-static analyses of basket support disks. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package performance under the Section 71.71 normal conditions of transport.
Page 14 of 37
2 38 Revise Tables 2.6.15.4-1, -2 and 2.6.15.5-1 of the SAR by also listing stress allowables
, .and corresponding design margins for the support disk.
Complete and accurate information should be presented in the SAR, per Section 71.7(a).
Section 2.7.1 Free-Drop (30-ft)-Cask Body Analysis 2-39 Identify the maximum design internal pressure used in determining cask bounding stresses.
Table 2.7.3.1-4 lists a maximum cask cavity pressure of 75 psig. Pg. 2.10.2-6 cites a pressure of 150 psig. On the basis of the Section 2.7.1 description, however, it is not clear whether a cask internal pressure of 75 psig is considered in the load combination evaluation for the 30-ft cask drops. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package performance under the Section 71.73 hypothetical accident conditions.
Section 2.7.1.5 Lead Slump Resulting From a Cask Drop Accident.
2-40 Submit an analysis of lead slump resulting from cask drop accidents.
. Supporting analyses are necessary to complete the review; the SAR provides only a summary description of lead slump evaluation results. Complete and accurate information should be provided, per Section 71.7(a).
Section 2.7.1.7 Closure Analysis 2 41 Submit an analysis of deformations and stresses for the protection lip of the cask top forging for the 30-foot free drop conditions.
The SAR uses the NUREG/CR-6007 approach for the closure analys,3, which provides that the closure bolts should be protected from direct impact to minimize bolt forces generated by free drops. The deformations of the protection lip should be shown to be less than the design diametric gap of 0.16 inch (78.36" - 78.20" = 0.16") between the closure lid and the protection lip under the free drop hypothetical accident conditions of 10 CFR 71.73 (c)(1).
Section 2.7.3 Thermal 2 The maximum fuel rod cladding temperature tabulated in Table 2.7.3.1-2 is inconsistent with that found in Chapter 3. Revise for consistency. Section 71.7(a) requires complete and accurate information.
2-43 Revise Table 2.7.3.13 to the corrected BWR canister internal pressure calculated in Section 3.5-8. Section 71.7(a) requires complete and accurate information.
Page 15 of 37
i Sections 2.7.8.4 & 2.7.10.4 FuelTube Analysis 2-44 Demonstrate the structural integrity of the fuel tube under the "line" load, as exerted by the spacer grid onto the mid-span of the fuel tube, in a cask side drop event.
The uniformly distributed " area" load may not yield bounding results, and loadings based on an equally realistic assumption of line load distribution should also be evaluated.
Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package under the free-drop condition and test of the Section 71.73(c)(1) hypothetical accident conditions.
Section 2.7.10 BWR Basket Analysis- Accident Conditions 2-45 Revise Tables 2.7.10.1-23 and -24 by also listing stress allowables and corresponding design margins for the support disk.
Complete and accurate information should be presented in the SAR, per Section 71.7(a).
2-46 Clarify, as appropriate, the underlined typographical or editorial errors. Complete and accurate information should be presented, per Section 71.7(a).
Table 2.7.10.1-24,"P m+ P Stresses for Support Disk... Thermal Case 2" Table 2.7.10.1-22, refers to thermal Case 4, in lieu of Case 2, for stress evaluation.
Section 2.10.1 Computer Program Description 2-47 Describe the revisions made to the RBCUBED program, subsequent to its previous application in July 1992, for the NAC Storable Transport Cask (NAC-STC).
The SAR refers to the November 1996 version of the program. However, it is not clear whether the 1992 version of the program has been modified and appropriately validated for the present Universal Transport Cask application. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the program for meeting the Section 71.101(a) quality assurance requirements.
Section 2.10.2 Finite Element Model-Universal Transport Cask 2-48 Justify the use of CONTACT 52 elements between the stacked annulus plates (item 33, Drawing No. 790-502) connecting the inner and t% outer shells of the cask.
The finite element analysis model should not allow force interaction between the annulus plates because a gap between the plates could potentially result from cask fabrication. Complete and accurate information should be provided, per Section 71.7(a),
for evaluating the adequacy of the cask finite element model.
Page 16 of 37
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2-49 Identify appropriately the cask internal pressures considered in evaluating cask
' structural pedormance under normal conditions of operation and hypothetical accident conditions.
Pg. 2.10.2-6 states, "A pressure of 159 psig is used to conservatively envelope the normal conditions design pressure of 25 psig for all impact loadings considered." Pgs.
2.6-4 and -14 cite a cask internal pressure of 50 psig for normal conditions of operation.
Pg. 2.7-1 discusses the application of the maximum design internal pressure to produce bounding stresses, but provides no pressure value for analyzing hypothetical accident conditions. Complete and accurate information should be provided, per Section 71.7(a),
for evaluating the package structural performance under the conditions and tests of Sections 71.71 and 71.73.
Section 2.10.3 Confirmatory Testing Program-UMS Impact Limiters and Attachments 2 50 Provide relevant test results and analyses to demonstrate that the RBCUBED program can be used to model the free-drop performance of the UMS Universal Transport Cask.
The SAR description suggests that the confirmatory testing program needs to be clarified for following inconsistencies in test execution and data reduction:
- 1. The 1/4-scale cask model should also provide proper simulation of the cask mass moment inertia, in addition to the mass and its center of gravity. The use of weight disks at only the cask top end may not be representative.
2, The measured acceleration time history in Figure 2.10.3-6 suggests significant cask rocking, which is uncharacteristic of a Universal Transport Cask undergoing side drop response.
- 3. The end drop acceleration time history in Figure 2.10.3-1 appears to contain much more spurious components than the similar time history for the 1/4-scale drop test conducted for the NAC-STC.
Complete and accurate information should be provided, per Section 71.7(a), for evaluating the testing program intended for confirming the calculated package performance under the free drops of Sections 71.71(c)(7) and 71.73(c)(1). :
1 2-51 Use all measured accelerometer time history traces for data evaluation and correlation with analytical results, and submit those traces and corresponding filtered results for staff review.
Measured accelerometer time histories from all four accelerometers should be considered to ensure that test results are consistently interpreted for data correlation evaluation. Complete and accurate information should be provided, per Section 71.7(a),
for evaluating the testing program intended for confirming the calculated package performance under the free drops of Sections 71.71(c)(7) and 71.73(c)(1).
2-52 Submit the force-deflection curves for the impact limiter models under the static crush ;
test configurations for the side and oblique drops. 1 l
Page 17 of 37
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In addition to the end drop static test results presented Figures 2.10.3-3 and -4,
. appropriate static test results should be shown to correlate adequately the RBCUBED calculated force-deflection curves of Figures 2.6.7.5-16 and -17, for the oblique and side drops, respectively. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the testing program intended for confirming the calculated package performance under the free drops of Sections 71.71(c)(7) and 71.73(c)(1).
2-53 Clarify, as appropriate, the underlined typographical or editorial errors.
Pg. 2.10.3-30, maximum accelerations summary in the data correlation table, "[U]pper impact limiter (peak positive or negative g values)... SH .. 49.57.. 50.M .."
The referenced SAR tables and figures suggest that some of the listed acceleration peak values are not related to the cask top-corner drop. Complete and accurate information should be presented, per Section 71.7(a).
Section 2.11 Site-Specific Contents Structural Evaluation l
2-54 Considering sectional (primary membrane and membrane-plus-bending), in lieu of ,
nodal, stresses in the support disk ligaments, re-evaluate normalized stress ratios in (
Table 2.11.1.1-1 for the Maine Yankee consolidated fuel. 1 The PWR support disk ligaments are evaluated with sectional stresses for the design basis spent fuel assemblies. When normalized stress ratios are considered in comparing relative structural performance, a consistent evaluation approach should be maintained throughout the SAR, including the Maine Yankee consolidated fuel. '
Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package structural performance under the conditions and tests of Sections 71.71 and 71.73.
2-55 Clarify the statement on Page 2.11.1-1, "[T]his study shows that a consolidated fuel assembly can be located in any position of the UMS PWR basket based on structural loading considerations."
Under a side drop, stresses in the support disk ligaments appear to be governed only by the locally applied equivalent inertia load of the design basis consolidated spent fuel assembly. As a result, because of the relatively large weight of the consolidated fuel lattice, some of the normalized stress ratios for the 12 fuel tube locations are expected to exceed 1.00, the stress ratio for Base Case. Complete and accurate information should be provided, per Section 71.7(a), for evaluating the package's structural performance under the conditions and tests of Sections 71.71 and 71.73.
2 56 Submit a stress summary table on maximum stresses in the support disk for location
" Case 6" to demonstrate adequate stress margins for the corner-location preferential loading of the consolidated fuel.
An evaluation of normalized stress ratios, in Table 2.11.1-1, alone may not be sufficient to substantiate the SAR conclusion on maximum stresses in the support disk, and explicit stress margins should be considered for the evaluation. Complete and accurate Page 18 of 37
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Information should be provided, per Section 71.7(a), for evaluating the package
. structural performance under the conditions and tests of Sections 71.71 and 71.73.
2-57 Clarify, as appropriate, the underlined typographical or editorial errors.
Pg. 2.11.2-1, "[T]he center of gravity for ...GTCC waster canister... identical to the C.G.
for the transport cask containing PWR Class 1 fuel (107.99 inches) as shown in Table 2.2-1."
Table 2.2-1 lists the location of C. G. at 106.60 inches from the bottom of the cask body; complete and accurate information should be provided, per Section 71.7(a).
I Page 19 of 37
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. CHAPTER 3 THERMAL Section 3.1 Discussion 3-1 Justify the use of a 10-year cooling time as a parameter of the design basis fuel.
The spent fuel contents in Chapter 1 have a minimum cool time of 6 years. The Thermal Section design basis fuel cool time of 10 years does not seem to bound the contents listed in Chapter 1. Section 71.7(a) requires complete and accurate information.
3-2 Provide justification for using a normal cladding temperature limit of 1058 'F, including supporting calculations from the methodology of the stated reference.
Typically, the cladding temperature limit of 1058 F is only for short-term accident conditions, short term off-normal conditions, or fuel transfer operations (like vacuum drying or dry transfer)- refer to NUREG-1536. Transportation of spent fuel, which can last for a period of 1 year, may not be a short-term event, and use of this higher temperature limit is not justified. The criticality analysis assumes that the fuel geometry remains intact, and the staff needs this information to complete the Section 71.51(a) and !
71.55 determinations.
Ths staff recognizes that the transportation regulations in 10 CFR Part 71 do not have any specific requirements for ensuring that the spent fuel cladding is maintained below its temperature limit. However, dual-purpose canisters must meet both transportation and storage requirements; and 10 CFR Part 72 requires that cladding be protected ,
during storage such that its degradation would not pose any operational safety problems - l with respect to its removal from storage (Section 72.122(h)(1)). Also, storage systems !
must be designed to allow ready retrieval of spent fuel for disposal ( Section 72.122(l)).
Therefore, for dual-purpose canisters one can readily deduce that cladding temperature limits that are imposed to prevent cladding damage during storage must also be met during transportation, especially if that fuel is to be stored post transport.
In addition, the guidance in Section 3.5.2.3 of the SRP (NUREG 1617), which is used to determine regulatory adequacy, requires that "... the maximum allowable fuel / cladding temperature is justified. The justification should consider the fuel and clad materials, irradiation conditions (e.g., the absorbed dose, neutron spectrum, and fuel burnup), and the shipping environment including the fill gas. Other necessary considerations include the elapsed time from removal of the spent nuclear fuel from the core to its placement into the transportation packaging, its time duration in the packaging, and its post-transport disposition."
3-3 Justify in the Thermal Section why the 16 kW and 20 kW limits bound the fuels to be transported. The justification should include considerations of the variable cooling times and enrichments. ,
This information is necessary to verify that the thermal analysis is bounding for the contents to be transported and the conditions specified in Sections 71.71 and 71.73.
Page 20 of 37
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Section 3.2.3 Convective Properties l
3-4 Provide verification of the properties and title of Table 3.2-4 as necessary to reflect the properties of Chemical Copper-grad Lead as opposed to chemicallead. Section 71.7(a) requires complete and accurate information.
Section 3.3.2 Safe Operating Ranges 3-5 Justify why the support disks are not within the safe operating temperature range and are not designated in Section 3.3.2 as a component that must be maintained within the i safe operating temperatup range.
For example, Table 3.4-1 lists the support disk interior temperature to be 686'F. Table 3.4-3 lists the support disk allowable range as -40 F to 650 F. Section 71.71(a) requires a determination of the effects on the design of the conditions and tests associated with the normal conditions of transport.
3-6 Correct typographical errors as needed for consistency between the table in this Section, and the reference included on pg. 4.5-7.
- I l
Pg. 4.5-7 lists the temperature range of the EPDM O-rings to be -65'F to 300*F. I Section 71.7(a) requires complete and accurate information.
3-7 Justify the ability of the neutron shield to perform its function when it exceeds its safe operating range during normal conditions of transport. Provide the shielding performance capability at the expected temperature range. )
At maximum PWR/BWR fuel decay and maximum ambient temperature, the temperature of the radial neutron shield exceeds the upper limit of the safe temperature range. Also, Section 3.5.2.3 of the SRP (NUREG-1617) states that ". . the temperature range of the thermal and structural properties for each package material exceed the specified and predicted temperature limits for the material." Section 71.71(a) requires a determination of the effects on the design of the conditions and tests associated with the normal conditions of transport. l 3-8 Justify the assumption that retaining the radial neutron shield during a 30-minute fire transient, and removing it afterward is the most conservative approach for determining component temperatures. Provide support via calculations.
The staff agrees that in a longer term fire transient, the approach may be conservative.
However, over a short period of time, the shield may act more as a barrier as the shield itself is heated. Section 71.41(a) requires an evaluation of the effects on the package of the tests specified in Section 71.73.
Section 3.4.1 Thermal Models 3-9 Justify neglecting the personnel barrier l'n toe models that arrive at component temperatures during the normal conditions of transport.
Page 21 of 37
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Section 3.5.4 of the Standard Review Plan suggests that the personnel barrier should
. .be considered when determining the package tempvatures for normal conditions of transport but should be neglected during the hypothetical accident.
Section 3.4.1.4 Personnel Barrier Thermal Model 3-10 include thermal radiation in the analysis of the accessible surface temperature.
In calculating the temperature of the personnel barrier, it is not conservative to neglect thermal radiation from the cask to the barrier. Section 3.5.4 of the Standard Review Plan suggests that the model consist of a heat balance at the surface of the package (at the personnel barrier) between the content decay heat and the convective and radiative heat losses to the environment at 100*F. Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
)
3-11 Specify the distance between the personnel barrier and the cask at the centerline of the )
model. Provide a figure similar to Figure 3.4-11 that includss temperatures at key points. Further, justify averaging the temperatures along the top of the personnel barrier to come up with a reported value which is only 3*F below the 122*F limit that the analysis in this Section is intended to meet.
The staff could not find a distance between the personnel barrier and the cask at the centerline. The method of averaging the temperatures does not seem justified if the personnel barrier is positioned close to the cask. It is not clear if the hottest region of the barrier would exceed the compliance criterion. Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
1 i
Section 3.4.4.1 Maximum Internal Pressure for PWR Fuel Canister and Cask 3-12 Review the normal conditions of transport and hypothetical accident conditions pressure calculations for appropriate term usage.
For example, pg. 3.4-28, should indicate 5,968 t/ canister. Another example, pg. 3.4 31, Vure Free Gas Volume calc, should be 1,096.25 t/ cask, as was appropriately used in the next equation determining the molar quantity of gas in the cask. These errors occur in both the PWR and BWR pressure calculations. Section 71.7(a) requires complete and accurate information.
Table 3.4-1 Summary Table of Temperatures - Maximum Component Temperatures-Normal Condition of Transport, Maximum Decay Heat, Maximum Ambient Temperature 3-13 Correct typographical errors as needed for consistency of temperatures between Tables 3.4-1, 2.6.1.1-1 and 3.4-3.
The guidance provided by the SRP suggests that the summary tables of temperatures of package components in the Thermal Section of the SAR must be consistent with the temperatures presented in the General Information Section and the Structural Page 22 of 37
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Evaluation Sections of the SAR for the normal conditions of transport and hypothetical
, accident conditions. Section 71.7(a) requires complete and accurate information.
Section 3.4.6 Evaluation of Package Performance for Normal Conditions of Transport -
3-14 Submit an analysis that justifies mixed loadings (i.e., spent fuel with longer cooling times and with lower temperature limits being loaded into a canister with shorter cooled spent fuel) are bounded. Indicate the effect the results have on the loading tables.
Since this transportation application includes spent fuel that is substantially different from the counterpart storage application (e.g., burnup of 50 vs 45 GWD/MTU) such a i comparison is warranted. 4 The current acceptable standard for establishing spent fuel cladding temperature lirris is PNL-6189 which determines a range of temperature limits depending on variations in fuel design, burnup level, cooling time, and storage cask design. Since multiple
{
temperature limits are likely to exist for a given dual-purpose canister design, it w suld be insightful to understand the possible limitations on loading the canister with fuel assemblies with different cladding temperature limits to ensure that the lowest cladding l temperature limit would not be exceeded. Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
Section 3.4.7 Thermal Evaluation - Direct Loading of the Universal Transport Cask 3-15 Demonstrate that the toughness of ASTM A69317-4 PH stainless steel, used for the PWR support disks, will be sufficient to withstand the hypothetical accident conditions aftir being exposed to temperatures near the prescribed short-term limit of 800*F during transfer operations. Clarify the temperature limits specified in Tables 3.4 3 and 3.4 9, and provide justifications for using those values.
As required in Section 71.43(f), the package must be designed, constructed, and prepared for transport so there will be no significant decrease in packaging effectiveness under normal conditions of transport. ASME Boiler and Pressure Vessel Code, Section ll, Part D, Table TM-1 indicates that this material may have reduced toughness at room temperature after being exposed to temperatures above 650*F.
Also, the value of the temperature limit for the 17-4 PH stainless steel support disks in Table 3.4-3 is inconsistent with the value in Table 3.4 9.
Section 3.5.1.1 Analytical Models 3-16 Clarify the method used to determine component temperatures during the hypothetical accident condition fire. Provide sufficient supporting calculation (s) to allow staff verifications of your results. Also, the explanation in the paragraph at the top of pg.
3.5-2 should be consistent with Footnote "b" of Tables 3.5-1 and 3.5-2.
Section 71.41(a) requires, in part, that the effects on a package of the hypothetical accident conditions must be evaluated by specific test or another acceptable method.
Page 23 of 37
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Clarification and supporting calculations are needed to allow the staff to independently
. verify your results.
Section 3.5.4.2.2 Maximum internal Pressure for Cask with BWR Fuel Canister (100%
Fuel Rod Failure) 3-17 Revise the pressure calculation of this Section, and enter the resulting pressure in Table 3.5-3 and consistently in other Chapters.
The cask volume as reported on pg. 3.4-43 (6895.73 liters / cask) should be used, not the canister volume. The staff recognizes that this error results in a conservative pressure.
Section 71.7(a) requires complete and accurate information. l i
Section 3.6.1.1 Maine Yankee Site-Specific Spent Fuel l 3-18 Consolidated Fuel
)
Justify the use of the 17x17 model used to obtain the effective conductivity of the l consolidated fuellattice with stainless steel rods at the perimeter. Re-assess the maximum fuel cladding temperature for all affected cases if needed.
The lattice model contains 41.5% stainless steel rods, while the actual case being analyzed is only 30% stainless steel dummy rods. This does not seem to be a i conservative representation of the effective conductivity of the consolidated lattice.
Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
3-19 Standard Fuel assemblies that have been repaired by removing damaged fuel rods and replacing them with stainless steel dummy rods, solidzirconium rods, or 1.95 wt %
enriched fuelrods Justify that cladding temperatures remain acceptable for replacement rods with low enrichment and variable burnup into a cask population of possibly cooler fuel, with lower l temperature limits.
Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
Page 24 of 37
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CHAPTER 4 CONTAINMENT Chapter 4.0 - Containment 41 Demonstrate that the design basis PWR and BWR fuels, as described in Section 4.0 of the containment analyses, bound the contents as described in Section 1.2.3. Revise Sections 4.0 and 1.2.3, as appropriate, to include all relevant points of this analysis. If the design basis fuels at described in Section 4.0 do not bound the contents as described in Section 1.2.3, recalculate the leakage rates for normal and hypothetical accident conditions. '
Section 71.33 requires a description of the package in sufficient detail to provide an adequate basis for its review. As indicated in Section 4.0, the design basis BWR and ;
PWR fuels are cooled 10 years at 50,000 MWD /MTU. However, Section 1.2.3 indicates that spent fuel contents have a maximum burnup of 50,000 mwd /MTU and a minimum cool time of 6 years.
4-2 Demonstrate that the design basis PWR and BWR fuels, as described in Section 4.0 of the containment analyses, bound all of the contents, including GTCC waste, that are described in Section 1.3.1. This can be accomplished by comparing the source terms and leakage rates of the contents in Section 1.3.1 to the source terms and leakage rates of the contents of the design basis fuel. If the design basis PWR and BWR fuels described in Section 4.0 do not bound the contents described in Section 1.3.1, recalculate the leakage rates for normal and hypothetical accident conditions. Revise Sections 4.0 and 1.2.3, as appropriate, to include all relevant points of this analysis.
Section 71.33 requires a description of the package in sufficient detail to provide an adequate basis for its review. There is insufficient technical basis to show that, from a containment analysis perspective, the contents described in Section 1.2.3 are bounded by design basis PWR and BWR fuels.
4-3 Update the SAR to reflect the correct value of the minimum enrichment for BWR fuel.
Section 71.7(a) requires complete and accurate information. Two different values for the minimum enrichment for BWR fuel (e.g.,1.9% and 3.25%) are specified in this Section.
Section 4.1 Containment Boundary 44 Revise the SAR to indicate that all leak tests will be performed in accordance with ANSI N14.5-1997. Describe how the confinement analyses, as presented in Chapter 4, comply.with the ANSI N14.5-1997 standard.
Section 71.31(c) requires identification of all established codes and standards applicable to the containment design. Currently, there is no reference to the ANSI N14.5-1997 '
standard in the Containment Section. This change is requested to update the SAR to current staff guidance that is specified in interim Staff Guidance No.11.
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I 45 Justify'why the cask lid bolts are considered part of the containment boundary.
Section 71.7(a) requires complete and accurate information. It is unclear why the cask lid bolts are considered to be part of the containment boundary.
Section 4.1.3.1 Seals 4-6 In a table, clarify the leak test criteria (i.e., air standard leak rate) and the test frequency for the containment system fabrication verification, periodic, maintenance, and pre-shipment leak tests. Also, in this table, indicate the corresponding replacement schedule for the o-rings. If there is a different allowable leakage rate for different contents (e.g., PWR fuel, BWR fuel, GTCC waste, or site-specific fuel), also note this in the table. Revise the appropriate Sections of Chapters 7 and 8 to clearly indicate the leak rate criteria for each type of test that will be performed and the seal replacement schedule.
Section 71.7(a) requires complete and accurate information. It is unclear what criteria are applied to each of the containment system fabrication verification, periodic, maintenance and pre-shipment leak tests. As an example of an area that needs clarification, Section 4.1.3.1.1 refers to the containment system fabrication verification ,
test as described in Section 8.1.3, and Section 4.1.3.1.2 refers to the leakage test !
procedures as described in Section 7.1.3. However, Section 8.1.3 does not clearly I describe any of the containment system leak tests, and Section 7.1.3 contains incorrect leak rates. I Section 4.2.1 Containment of Radioactive Material 4-7 include the values of the following parameters used in the containment analysis leakage rate calculations for normal conditions of transport: fraction of rods that develop breaches (f ), crud surface activity (Sc), free volume inside the containment vessel (V),
capillary length (a), the hole diameter (D), and the gas temperature (T) and upstream pressure (Po ) that were used to calculate the hole diamater. Include values for both PWR and BWR fuels, as appropriate.
Section 71.7(a) requires complete and accurate information. The values for these parameters are not presented. The staff needs the information to perform confirmatory i analysis and to verify that bounding values were used for the containment analysis, i l
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48 include the calculations that were used to determine the leak rates and sensitivities for the PWR and BWR baskets under helium leak testing conditions. Also, list the value of the parameters used as inputs to the calculations.
Section 71.7(a) requires that the SAR contain complete and accurate information. Only the leak rate and sensitivities are included as a note to Table 4.2-4. There is not enough information to verify that these leakage rates are bounding for test conditions.
The staff needs the information to perform confirmatory analysis and to verify that bounding values were used for the containment analysis.
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Section 4.5.3 SAS2H Output and Group A2Values for B&W 15x15 and GE 9x9 Assemblies 4-9 Include the SAS2H input files for PWR and BWR design basis fuel source terms.
Section 71.7(a) requires complete and accurate information. The input files for PWR and BWR design basis fuel source terms are not presented. The staff needs the input i files to verify that the design basis fuels represent the bounding source terms for the l containment analysis.
4 10 in Table 4.5.3-5, correct the A2 values, and recalculate the leakage rates, for transport casks containing BWR fuel.
Section 71.7(a) requires complete and accurate information. The incorrect A2 values were used to calculate the Group A2 for volatiles.
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CHAPTER 5 SHIELDING Section 5.1.3 Results of Analysis 5-1 For Tables 5.1-1,5.1-2,5.1-3,5.1-4,"PWR and BWR Maximum total dose rate summaries for normal and accident conditions": provide a figure identifying the locations of the maximum dose rate values. The staff needs the information to confirm that the application. meets the external radiation standards of Section 71.47.
Section 5.2 Source Specification 5-2 Provide a list of resources that confirm justifying the Co-59 impurity level of 1.2 g/kg in the Type 304 stainless steel hardware.
Section 71.33 requires the identification of the maximum radioactivity of the package constituents. This information is necessary to confirm the validity of your results.
5-3 For Tables 5.2-10 and 5.2-11. "One-Dimensional Dose Rate Results Relative to PWR and BWR, respectively, Design Basis Fuel": For the normal conditions, radial dose rates are for surface and 2.4 meters. Provide an explanation of why a distance of 2.4 meters was used. The staff needs the information to confirm that the application meets the external radiation standards of Section 71.47.
Section 5.3 Model Specifications 5-4 Provide figures that clearly distinguish the regions and materials for the following figures, and provide the names of the materials listed in the legend. Section 71.33 requires a description in sufficient detail to identify the package accurately and provice a sufficient basis for evaluation.
Figure 5.3-14 PICTURE Representation of PWR Top Model- Normal Conditions
- Showing Trunnion Recesses and Lid Vent Figure 5.3-15 PICTURE Representation of PWR Bottom Model- Normal Conditions - Slice Through Lower Rotation Pockets Figure 5.3-17 PICTURE Representation of PWR Top Model- Accident Conditions Figure 5.3-18 PICTURE Representation of BWR Fuel Region and Heat Transfer Models at Fuel Axial Midplane Section 5.3.1.1 One-Dimensional Radial Model 55 The paragraph references several figures, however, Figure numbers have been omitted from the text. Section 71.7(a) requires complete and accurate information.
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Section 5.4.2.1 Normal Conditions of Transport 5-6 . Provide additional explanation of the results depicted in the graph in Figure 5.4-8. I The effect of the heat transfer fins on the dose rate is shown in Figure 5.4 8. Justify the assumption that the localized peak in neutron dose rate is offset by a localized depression in the gamma contribution. Also, explain the apparent lack of symmetry to the projected dose rates in the graph. In addition, revise Figure 5.3.3 to show the heat transfer fins.
I Section 71.33 requires a description in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation.
l Section 5.4.3 Loading Table Analysis 5-7 For Table 5.4-22, " Loading Table for BWR Fuel": clarify if low enrichment fuel (1.9%- l 2.3%) with 45<Burnup<50 GWD/MTU is expected, or clearly specify that it has not been 4 evaluated.
The high burnup, low enriched fuel requires a cooling time greater than 40 years. It does, however, appear to be included in the loading table. This is a bounding fuel and must be evaluated if it is intended to be placed in the Universal Transport Cask. Section 71.7(a) requires complete and accurate information.
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CHAPTER 6 CRITICALITY Section 6.1 Discussion and Results 6-1 Clarify the conflicting statements on the maximum initial enrichment for BWR fuel contents.
The maximum initial enrichment of BWR fuel is stated at various places in the SAR to be either 3.75 or 4.0 wt% 835U. A maximum enrichment of 4.0 wt% 5U appears in Sections 1.2.3,6.3.2,6.3.3,6.4.1.3.2, and 6.4.3.1, whereas 3.75 wt% 23sU appears in Sections 6.1 and 6.3.4.1 and in the CSAS computational inputs shown in Figures 6.6.2-3 and 6.6.2-4.
Section 71.33(b) requires an accurate description of the package contents. SRP Section 6.5.2 identifies initial enrichment as an important specification for spent fuel contents.
Section 6.2 Package Fuel Loading 6-2 For each of the five fuel / canister classes, describe how axial poison coverage is ensured i under the hypothetical accident conditions of transport. I The staff's preliminary calculations indicate that, following an end-drop accident, the active fuel tips can protrude significantly beyond the ends of the UMS poison panels for certain fuel designs with especially short dimensions of the top hardware. Therefore, in showing that all allowed fuel contents are within the analyzed safety basis, NAC should !
describe in detail how axial poison coverage is determined. The determination should ;
consider the extent to which a top or bottom end-drop accident could axially dislocate the fuel pins within the assemblies and the active fuel within the fuel pins. Conclusions regarding the potential for end-drop accidents to produce hardware damage and material dislocations in the fuel assemblies should also be reflected in the description of modeling assumptions in Section 6.3.2.
Section 71.55 requires evaluation of the most reactive credible configurations. SRP Section 6.5.1.1 specifies the review of features that locate the fuel relative to neutron absorbing material. SRP Section 6.5.3.1 specifies considering off-nominal relative positionings of fuel and basket components.
Section 6.3.4.1 Fuel Region 6-3 Correct the inconsistency in the tabulation of fuel material densities.
The second material in the left column is UO, with 3.75 wt% 8 5U, yet the atom densities i listed in the right column correspond to UO, with 4.0 wt% rasU. The latter materialis inconsistent with the CSAS inputs and outputs shown in Figures 6.6.2-3 and 6.6.2-4.
Section 71.33(b) requires an accurate description of the package contents. SRP Section ;
6.5.3.2 specifies verification of appropriate mass densities and atom densities for all modeled materials of the packaging and contents.
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Section 6.3.4.2 Cask Material, 6-4 Either correct the "BORAL core" compositions used in the CSAS models or provide a calculation showing that the as modeled compositions do not lead to significant errors in the calculated values of k,n.
The staff has confirmed that the "BORAL core" material densities used in NAC's CSAS models do lead to the appropriate areal densities of ' B. However, the model's densities of "B, C, and Al are significantly off. For example, '0B comprises 15.4% instead of 20%
of the boron atoms, and the stoichiometric ratio of boron to carbon atoms for B4 C is 38 (tabulated) or 3.8 (CSAS output) instead of 4.
Section 71.33(b) requires an accurate description of the package materials used for neutron absorption. SRP Section 6.5.3.2 specifies verification of appropriate mass densities and atom densities for all modeled materials of the packaging and contents. -
Section 6.4.3.1 Summary of Maximum Criticality Values 6-5 Correct "75% of nominal -. . . " to "75% of minimum ' B loading." -
The staff notes that other Sections of the SAR refer to the minimum ' B loading.
SRP Section 6.5.3.2 states that no more than 75% of the specified minimum neutron poison concentration of the packaging should be considered in the criticality evaluation.
Any deviations from the SRP guidance should be justified.
Table 6.5-3 : Most Reactive Configuration System Parameters 6-6 Correct the enrichment entry in this table from 4.0 to 4.2 wt% "U.
Section 71.33(b) requires accurate description of the package contents. SRP Section 6.5.2 identifies initial enrichment as an important specification for spent fuel contents.
Appendix 6.6.1.1 Criticality Evaluation of Maine Yankee Site-Specific Spent Fuel 6-7 . Describe the applicability of the critical benchmark experiments to the site specific Maine Yankee contents.
The current discussion in Section 6.5.1.2 covers the applicability of experiments to the standard fuel contents only. A similar discussion for Maine Yankee contents does not appear in this appendix for site-specific spent fuel. The requested description should compare the parameters H/U volume ratio and the average group causing fission for the most reactive site-specific contents to those for the experiments.
Section 71.41 requires that compliance be demonstrated by methods acceptable to the Commission. SRP Section 6.5.7.1 states that the applicant should justify that the benchmark experiments are applicable to the package design and contents.
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CHAPTER 7 OPERATING PROCEDURES Chapter 7.0 Operating Procedures 7-1 in Table 71, correct the torque values and corresponding tolerances in accordance with the License Drawings for the following components:
Comoonent Toraue Value (Ch. 7) Toraue Value on License Drawino Coverplates 300120 ft lbs 300150 in-Ib (Dwg. No.790 500) 300*20 in-lb (Dwg No. 790-503)
Secondary Trunnion Bolts 500i10 ft-Ib 500*50 ft lb (Dwg. No. 790-500)
Section 71.7(a) requires complete and accurate information. Update the SAR Drawings or Chapter 7, as appropriate, to reflect the correct torque values.
72 In Table 7-1, clarify which component the " test plug" is associated with, and add a more descriptive entry for this component to the table. -
Section 71.7(a) requires complete and accurate information. There are three test ports on the Universal Transport Cask used for helium leak testing. One is used to test the cask lid o-ring, while the other two are used to test the o-rings of the two port coverplates. The description of the test plug in Table 7-1 does not distinguish which test plug is torqued to 3013 ft-Ib.
7-3 Describe procedures for conducting the helium leak tests of the lid and port coverplates.
The procedure should also include a discussion of the corrective actions if the components or configurations fail to meet the test criteria.
Section 71.7(a) requires complete and accurate information, and Section 71.87 requires that the gaskets be properly installed and secured and free of defects. The application does not contain enough of a description to ensure the adequate conduct of the helium leak test to assure that the o-rings are properly installed.
7-4 Include procedures for the preparation of the Universal Transport Cask for wet loading, and describe the basis for wet loading the TSC or the GTCC waste canister into the cask.
Section 71.87(f) requires that a verification must be made to ensure that the package has been loaded and closed appropriately. Section 7.5.1 discusses loading fuelinto the TSC while it is installed in the Universal Transport Cask. This is referred to as wet loading. However, there are no procedures for preparing the Universal Transport Cask for the wet loading operation, and there is no discussion of the purpose for wet loading of the fuelin the SAR.
7-5 Revise the operating procedures to include appropriate controls for detecting and preventing the ignition of any combustible gases during welding, grinding, or cutting l operations associated with closure of the TSC or the GTCC waste canister. j Page 32 of 37
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Section 71.43(d) requires that a package be made of materials that assure there will be
'no significant chemical or galvanic reactions. Potential reactions between the TSC or Universal Transport Cask paint coatings and/or other components (such as the aluminum heat transfer disks) with spent fuel pool water may produce hydrogen or other flammable gases. Since the shield lids of the TSC and GTCC waste canister are ;
welded to their shells during fuel loading, there is a source of heat that could lead to ignition if sufficient amounts of gas are present. 1 Section 7.3 Recieving Universal Transport Cask and Unloading Transportable Storage Canister From Universal Transport Cask i 7-6 Revise the SAR to include operating procedures for unloading the contents of the TSC and GTCC waste canisters. Section 71.7(a) requires complete and accurate information, l
Section 7.5.1 Loading and Closing the Transportable Storage Canister Containing J Spent Fuel I 7-7 Specify the minimum helium gas purity that will be used to vacuum dry and backfill the cask cavity, TSC, and the GTCC waste canister. Perform a calculation to show that this minimum helium gas purity assures that the spent fuel will not undergo significant degradation during transport.
Section 71.55(d)(2) requires that the geometric form of the package contents of a fissile materials package will not be substantially alk ad under the test conditions specified for normal conditions of transport. The SAR dot.., not specify the minimum helium gas purity that is used to vacuum dry and backfill the cask cavity, TSC, and GTCC waste canister. PNL-6365," Evaluation of Cover Gas impurities and Their Effects on the Dry i Storage of LWR Spent Fuel," recommends that the amount of oxidizing gases in a '
storage cask be limited to 1 gram-mole. This 1 gram-m, ole limit reduces the amount of oxidants below levels where any cladding degradation is expected.
7-8 Revise the operating procedure to require a TSC vacuum hold time of 30 minutes, or provide a justification for using 20 minutes.
Section 71.55(d)(2) requires that the geometric form of the package contents of a spent fuel package will not be substantially altered under the test conditions specified for normal conditions of transport. The procedure in the SAR specifies a hold time of 20 minutes. However, PNL-6365," Evaluation of Cover Gas impurities and Their Effects on the Dry Storage of LWR Spent Fuel," recommends that a constant pressure of 3 Torr should be maintained for 30 minutes without the aid of a vacuum pump to assure that oxioizing gases are sufficiently removed from the cask to prevent gross degradation of i the cladding due to excessive fuel oxidation.
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7-9 Revise the allowable leakage rate and sensitivity to reflect the correct values as !
specified in Chapter 4. In the procedure, also include the allowable leakage rates for canisters containing both PWR and BWR fuels.
Section 71.7(a) requires complete and accurate information. Incorrect values for the '
leakage rate and sensitivity are specified in the operating procedure.
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.,e CHAPTER 8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Section 8.1.3 Leak Tests 8-1 Revise the SAR to include the following details:
- 1. a reference to your commitment to use ANSI N14.5-1997 as the accepted standard for leak testing;
- 2. the allowable leakage rates for the Universal Transport Cask to be consistent with the containment analysis of Chapter 4 for both PWR and BWR fuels;
- 3. descriptions of the methods used to assure that the test plugs of the Universal Transport Cask lid and the two port coverplates are adequately sealed; and
- 4. description of the helium leak test acceptance or rejection criteria.
Section 71.7(a) requires complete and accurate information. There is not enough information in Chapter 8 to conclude that the helium leak testing will be performed in accordance with the specifications of ANSI N14.5-1997.
8-2 Clarify the reference made to the fabrication leak tests. -
Section 71.7 requires complete and accurate information. A reference is made to the fabrication leak tests of Section 8.1.3. However, Section ( 1.3 does not describe the leak tests acceptance criteria applied to the containment ' tem fabrication verification, periodic, maintenance or the pre-shipment leak tests.
Section 8.1.7 Neutron Absorber Verification Tests 8-3 Describe the reference BORAL standard plates and the testing that is required for qualifying them as reference standards.
The description of the BORAL standard plates used for calibration should address the size distributions of B,C particles, the absolute magnitude and uniformity of the ' B areal densities, and how these characteristics are determined. To bound any effects of j neutron channeling on the calibration, the ' B isotopic composition (natural or enriched) l should be no higher and the B4C particles should be no larger in the standards than in the panels. If neutron transmission or radiography measurements are used for qualifying the reference BORAL standard plates, the calibration standards used for those measurements should likewise be described.
Section 71.123 and 71.125 require the establishment of a program and measures to identify, perform, and control the testing required to demonstrate that the packaging components will perform satisfactorily in service. SRP Section 6.5.3.2 calls for ensuring that the neutron absorbers are properly controlled during fabrication to meet their specified properties.
8-4 Characterize (a) the uncertainty in the neutron luminance measurement and (b) the sensitivity of the measured luminance to ' B content over the two ranges of areal density,0.011 and 0.025 g/cm 2, for the BWR and PWR BORAL panels, respectively.
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Section 71.123 and 71.125 require the establishment of a program and measures to identify, perform, and control the testing required to demonstrate that the packaging components will perform satisfactorily in service. SRP Section 6.5.3.2 calls for ensuring that the neutron absorbers are properly controlled during fabrication to meet their specified properties.
Section 8.2.1 Structural and Pressure Tests 8-5 Address the annual testing provisions, per your commitments to meet ANSI N14.6.
The SAR discusses only visual inspection of trunnions prior to each shipment. Sections 6.3 and 6.5 of ANSI N14.6, however, provides that a liquid penetrant or magnetic particle examination shall be performed for the trunnions, if a load testing is omitted.
Section 8.3.2 Cask Body Fabrication j
8-6 Include a more clear description of the leak test criteria (i.e., air standard leak rate) and l the test frequency for the containment system fabrication verification, periodic, l maintenance, and pre shipment leak tests. -
Section 71.7(a) requires complete and accurate information. Section 8.1.3 does not clearly describe the criteria applied to each of the containment system fabrication verification, periodic, maintenance, and pre-shipment leak tests.
Section 8.3.3.2 Lead Pour Operations 8-7 Specify the correct grade of lead that will be poured into the cask annulus.
Section 71.7(a) requires complete and accurate information. This Section does not indicate which grade of lead will be poured into the cask anrolus. Table 2.3.7-1 and Drawing No. 790 502 indicate that " Chemical-Copper Grade" lead will be used.
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Dual Purpose Canister During the course of its review, the staff identified various issues that, while not directly relevant to transportation, may be applicable to the intended storage of the transported contents. These issues are not required to be addressed to obtain a 10 CFR Part 71 Certificate of Compliance.
DP4-1 Specify, and provide the technical basis and calculations for, the allowable leakage rate and sensitivity of the TSC and the GTCC waste canister. Revise Sections 7and 8 as appropriate to include the allowable leakage rate'and sensitivity for the TSC and GTCC waste canister.
Dual-purpose canisters must meet both transportation and storage requirements for use under the respective regulations. Sections 72.24(d),72.104(a), and 72.106(b) require that the storage system will reasonably maintain confinement of radioactive materials under Section Part 72 normal, off-normal, and credible accident conditions. Since the j TSC and the GTCC waste canister may be used for storage following transport, an '
analysis of the TSC and GTCC waste canister leakage rates and sensitivities is needed to assure that the canisters perform their intended function during storage.
DP4-2 Include a discussion of the welding requirements of the TSC and the GTCC waste canister.
Dual-purpose canisters must meet both transportation and storage requirements for use under the respective regulations. Section 72.122(l) requires that the storage system be designed to allow ready retrieval of spent fuel for further processing or disposal. Since j the TSC and the GTCC waste canister are integral components of the Universal j Transport Cask, a description of the welding requirements for the canisters is needed to j assure that the canister can be removed from the Universal Transport Cask.
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DP7-1 For the TSC and the GTCC waste canist9r, demonstrate that the 3/8 inch maximum welding pass thickness and associated nondestructive examination methods will provide reasonable assurance that the confinement system maintains structural integrity for the situation where the canister is placed into a storage overpack for long-term storage.
The discussion should address the points of Interim Staff Guidance No. 4. Update the appropriate Sections (e.g., Sections 4.1.3.2, etc.) to address the welding procedures and acceptance criteria.
Dual-purpose canisters must meet both transportation and storage requirements for use under the respective regulations. Section 72.122(l) requires that the storage system must be designed to allow ready retrieval of spent fuel for further processing or disposal.
Since the TSC and GTCC waste containers are integral components of the Universal Transport Cask, a description of the welding requirements for the canister is needed to assure that the canister can be removed from the Universal Transport Cask.
DP8-1 include the allowable leakage rate and sensitivity for the TSC and the GTCC waste canister.
Dual-purpose canisters must meet both transportation and storage requirements for use under the respective regulations. Sections 72.24(d),72.104(a), and 72.106(b) require that the storage system will reasonably maintain confinement of radioactive materials Page 36 of 37
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,- t ,o e under Section Part 72 normal, off-normal, and credible accident conditions. Since the TSC and the GTCC waste canister may be used for storage following transport, an analysis of their leakage rates and sensitivities is needed to describe the canisters intended functions during storage.
DP8-2 include the acceptance or rejection criteria and describe the repair methods for welding the TSC and GTCC waste canister lids onto the shells.
Dual-purpose canisters must meet both transportation and storage requirements for use .
under the respective regulations. Section 72.122(l) requires that the storage system must be designed to allow ready retrieval of spent fuel for further processing or disposal.
Since the TSC and GTCC waste canister are integral components of the Universal Transport Cask, a description of the welding requirements for the canisters is needed to assure that the canisters can be removed from the Universal Transport Cask.
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