ML18355A831

Tn Americas, LLC - MP197 Transportation Packaging Safety Analysis Report, Revision 18D, Changed Pages Only. (Non-Proprietary)

ML18355A831
Person / Time
Site:	07109302
Issue date:	12/31/2018
From:	Orano USA, TN Americas LLC
To:	Office of Nuclear Material Safety and Safeguards
References
E-52844, EPID-L-2018-LLA-0000
Download: ML18355A831 (84)

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{{#Wiki_filter:Enclosure 8 to E-52844 NUHOMS-MP197 SAR, Revision 18D, Changed Pages Only (Non-Proprietary)

MPl 97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Revision Log Rev. No. Date Description 0 4/2001 Original Issue for CoC Revision 0 1 1/2002 Various Changes CoC Revision 0 2 2/2002 Various Changes CoC Revision 0 3 4/2002 Various Changes CoC Revision 0 4 5/2002 Various Changes CoC Revision 0 5 3/2009 Application for CoC Revision 3 Added Appendix A for the MP197HB 6 6/2009 Various Changes for CoC Revision 3 7 4/2010 Various Changes for CoC Revision 3 8 7/2010 Various Changes for CoC Revision 3 9 3/2011 Various Changes for CoC Revision 3 10 8/2011 Consolidated SAR Submittal for CoC Revision 3 11 9/2011 Various Change§ for CoC Revision 4 12 2/2012 Application for CoC Revision 7 13 8/2012 Various Changes for CoC Revision 7 14 8/2013 Various Changes for CoC Revision 7 15 1/2014 Various Changes for CoC Revision 7 16 3/2014 Various Changes for CoC Revision 7 17 4/2014 Various Changes for CoC Revision 7 18 4/2017 Application for CoC Revision 8 Revised pages as follows: SAR pages A.l.4.8-3a, A.l.4.8.13a, A.1.4.8-14, A.1.4.9-i, A.1.4.9-2, A.l.4.9-2a, A.1.4.9-3, A.1.4.9-9, A.1.4.9-9a, A.1.4.9-10, A.1.4.10-1, A.1.4.10-6 SAR pages A.2.13.11-ii, A.2.13.11-1 through A.2.13.11-6, A.2.13.11-12, A.2.13.11-13, A.2.13.11-18, A.2.13.11-22 through A.2.13.11-24, A.2.13.11-31, A.2.13.11-41, A.2.13.11-43 SAR pages A.3-i through A.3-v, A.3-1, A.3-3, A.3-4, A.3-6, A.3-7, A.3-28, A.3-55, A.3-62, A.3-77, A.3-78, A.3-80, A.3-81, A.3-97, A.3-98, A.3-99, A.3-102, A.3-175, A.3-176 SAR pages A.5-i through A.5-v, A.5-1, A.5-2, A.5-3, A.5-5, A.5-7, A.5-8, A.5-11, A.5-12d, A.5-16, A.5-16a, A.5-16b, A.5-20b, A.5-29a, A.5.-29c, A.5-29d through A.5-29h, A.5-32b, A.5-34, A.5-37, A.5-38, A.5-40, A.5-41, A.5-41a, A.5-70, A.5-75, A.5-801, A.5-80m, A.5-80p, A.5-80q, A.5-80t, A.5-105 NUH09.0101 1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Revision Log Rev. No. Date Description Revised drawings as follows: Drawing MP197HB-71-1005 Drawing NUH69BTH-71-1004 New pages as follows: Proprietary Information Notice Revision Log Appendix A Master Table of Contents, pages i through xix SAR pages A.1.4.9-9b through A.1.4.9-9g, A.1.4.9-1 Sa SAR pages A.3-60a, A.3-102a, A.3-102b SAR pages A.5-2a, A.5-29i, A.5-38a, A.5-41b, A.5-4lc, A.5-42a, A.5-79b, A.5-102a 18A 2/2018 Application for Revision 9 Revised pages as follows: SAR pages A.1-2, A.1-4, A.1.4.9Ai, A.1.4.9A-1 throughA.1.4.9A-4, A.1.4.10-i, A.1.4.10-1, A.1.4.10-6, andA.1.4-10-307 A.2-3, A.2-6, A.2-7, A.2-16, A.2-23, A.2.13.1-3, A.2.13.2-1, A.2.13.2-4, A.2.13.2-13, A.2.13.2-15, A.2.13.2-18, A.2.13.2-30, A.2.13.2-31, A.2.13.5-18, A.2.13.12-10, A.2.13.12-18, and A.2.13.14-5 A.4-i, A.4-1, A.4-3, A.4-4, A.4-5, and A.4-6 A. 7i, A. 7.ii, A. 7-2, A. 7-3, A. 7-4, A. 7-5b, A. 7-6, A. 7-7, A. 7-8, A. 7-8a, A.7-9 throughA.7-14, A.7-17, A.7-21, A.7.7.10.i, andA.7.7.10-1 through A. 7. 7.10-3 A.8-i, A.8-3, A.8-15, A.8-16, and A.8-17 Revised drawings as follows: Drawings MP197HB-71-JOOJ through MP197HB-71-1009, MP197HB-71-JOJJ MP 197HB-71-1014 NUHRWC-71-1001 NUH09.0101 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Revision Log Rev. No. Date Description New pages as follows: A.1.4.9A-5 A.4. 6.1-i and A.4. 6.1-1 through A.4. 6.1-11 A. 7-3a, A. 7-4a, and A. 7-6a A.8-15a 18B 6/18 Application for CoC Revision 9 Revised pages as follows: SARpages 1-i, 1-13, 1-14, 1-15, 1-16 SAR pages A.1.4.10-1, A.1.4.10-6 SAR pages A.2-13, A.2.13.7-i, A.2.13.7-40, A.2.13.7-44 SARpage 7-18 Revised drawings as follows: MP 197HB-71-1004 through MP 197HB-71-1006 MP197HB-71-1009 NUHRWC-71-1001 18C 8/18 Application for CoC Revision 9 Revised pages as follows: SAR page A.1.4.10-1 SAR page A. 7-18 Revised drawings as follows: MP 197HB-71-1014 NUH09.0101 3

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Revision Log Rev. No. Date Descriotion 18D 10/18 Application/or CoC Revision 9 Revised pages as follows: SAR pages A.1-3, A.1-8, and A.1-9 SAR pages A.1.4.9A-1, A.1.4.9A-2, A.1.4.9A-3, A.1.4.9A-4, and A.1.4.9A-5 SAR pages A.1.4.10-i, A.1.4.10-1, andA.1.4.10-6 SAR pages A.2.13.5-20, A.2.13.5-22, A.2.13.5-26, andA.2.13.5-29 SAR pages A.2.13. 7-i, A.2.13.7-ii, A.2.13. 7-1, A.2.13. 7-4, A.2.13. 7-5, A.2.13. 7-6, A.2.13. 7-7, A.2.13. 7-8, A.2.13. 7-9, A.2.13. 7-10, A.2.13. 7-11, A.2.13. 7-13, A.2.13. 7-14, A.2.13. 7-43, A.2.13.7-44, A.2.13.7-48, andA.2.13.7-49 SAR pages A.2.13.9-2, A.2.13.9-4, A.2.13.9-6,\\A.2.13.9-7, A.2.13.9-9, andA.2.13.9-12 SAR pages A.2. 13.14-7 SAR pages A.5-i, A.5-iv, A.5-3, and A.5-6 SAR pages A. 7-i, A. 7-1a, A. 7-17, and A. 7-18 SARpagesA.8-4 andA.8.5 Revised drawings as follows: Drawings MP 197HB-71-1001, MP 197HB-71-1002, MP 197HB-71-. 1004 throughMP197HB-71-1006, MP197HB-71-1008, MP197HB-71-1009, and NUHRWC-71-1001 New pages as follows: Revision Log Page 4 A.1.4.9A-3a, A.2.13. 7-3a, A.2.13. 7-4a, A.2.13. 7-6a, A.2.13. 7-8a, A.2.13. 7-43a, A.2.13. 7-43b, A.2.13. 7-43c, A.2.13. 7-48a, A.2.13.7-49a, andA.2.13.7-57a A.5-1b, A.5-1c, A.5-3a, A.5-6a, A.5-6b, A.5-38a, A.5-38b, A.5-80u, A.5-80v, and A.5-80w A. 7-18a, A. 7-18b, and A. 7-18c A.8-4a NUH09.0I01 4

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A personnel barrier is mounted to the transport frame to prevent unauthorized access to the cask body. There are nine DSC designs that constitute a portion of the NUHOMS-MP197HB packaging. All of the DSCs consist of a cylindrical shell, and top and bottom shielded closure assemblies. Details for each DSC type are provided in Appendices A.1.4.1 through A.1.4.9. After loading, each DSC is vacuum dried and back-filled with an inert gas. Each DSC includes a fuel basket assembly, located inside the DSC. The basket assembly locates and supports the fuel assemblies, transfers heat to the DSC wall, and provides neutron absorption to satisfy nuclear criticality requirements. For some DSC designs, a basket hold down ring is installed on top of the basket, after fuel loading, to prevent axial motion of the basket within the DSC. The dry irradiated and/or contaminated non-fuel bearing solid materials are contained in a secondary container (Radioactive Waste Canister (RWC)). The safety analysis of this configuration takes no credit for the containment provided by the RWC. A.1.2.1.1 NUHOMS-MP197HB Transport Cask The cask is fabricated primarily of nickel-alloy steel. (NAS). Other materials include the cast lead shielding between the containment boundary inner shell and the structural shell, the 0-ring seals, the borated resin neutron shield and the carbon steel closure bolts. Socket headed cap screws (bolts) are used to secure the lid to the cask body and the ram access closure plate to the bottom of the cask. The body of the cask consists of a 1.25 inch thick, 70.50 inch inside diameter NAS inner (containment) shell and a 2.75 inch thick, 84.50 inch outside diameter NAS structural shell which sandwich the 3.00 inch thick cast lead shielding material. MP 197HB Unit 01 as-fabricated cast body has a reduction in its shielding capability due to localized areas where the lead thickness is less than 3.00 inches. See Appendix A.1.4.9A, Chapter A. 5 and Chapter A. 7 for further details. The overall dimensions of the NUHOMS-MP197HB packaging are 271.25 inches long and 126.00 inches in diameter with both impact limiters installed. The transport cask body is 210.25 inches long and 84.50 inches in diameter. The cask diameter including the radial neutron shield is 97.75 inches or 104.25 inches with the fins. The minimum length of cask cavity is 199.25 inches and 70.50 inches in diameter without the sleeve or 68 inches with the sleeve. Detailed design drawings for the NUHOMS-MPl 97HB packaging are provided in Appendix A.1.4.10, Section A.1.4.10.1. The materials used to fabricate the packaging are shown in the Parts List on Drawing MP197HB-71-1002. Where more than one material has been specified for a component, the most limiting properties are used in the analyses in the subsequent sections of this appendix to the SAR. The maximum gross weight of the loaded package is 152.0 tons including a maximum payload of 56.0 tons. Table A.1-1 summarizes the dimensions and weights of the NUHOMS-MP197HB packaging components. Trunnions, attached to the cask body, are provided for lifting and handling operations, including rotation of the packaging between the horizontal and vertical orientations. The NUHOMS-MP197HB packaging is transported in the horizontal orientation, on a specially designed shipping frame, with the lid end facing the direction of travel. DSCs with a spent fuel payload are shipped dry in a helium atmosphere. Both the transport cask cavity and the DSC cavity are filled with helium. The heat generated by the spent fuel assemblies NUH09.0101 A.1-3 All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.1-1 Nominal Dimensions and Weights of the NUHOMS-MP197HB Packaging Nominal Dimensions (in.) NUHOMS-MP197HB packaging overall length with impact limiters and thermal shield 271.25 NUHOMS-MPI97HB packaging overall length without impact limiters and thermal shield 210.25 NUHOMS-MP197HB cask impact limiter outside diameter 126.00 NUHOMS-MP197HB cask outside diameter (w/o impact limiters and fins) 97.75 NUHOMS-MP197HB cask outside diameter with fins (w/o impact limiters) 104.25 NUHOMS-MP197HB cask cavity inner diameter 70.50 NUHOMS-MP197HB cask cavity length (minimum) 199.25 NUHOMS-MP197HB cask inner shell radial thickness 1.25 NUHOMS-MP197HB cask lead gamma shield radial thickness 3.00(1J NUHOMS-MP197HB cask body outer shell 2.75 NUHOMS-MP197HB cask lid thickness 4.50 NUHOMS-MP197HB cask bottom thickness 6.50 NUHOMS-MP197HB cask resin and aluminum box thickness 6.25 Nominal Weights (lb x 1000) Weight of Contents (maximum) 112.0 Empty weight ofNUHOMS-MP197HB Packaging without lid or impact limiters 157.5 Cask lid 6.0 Outer sleeve with fins 3.1 Weight of impact limiters, thermal shield, and attachments 25.0 Total loaded weight ofNUHOMS-MP197HB Packaging (without transport skid) 303.6 (I) MP 197HB Unit OJ has localized areas where the lead thickness is below 3.00 inches. NUH09.0101 A.1-8 All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev.18D, 12/18 I Table A.1-2 DSC Configuration in the NUHOMS-MP197HB Package Sub Bottom Sleeve Fins Detailed Contents DSC Type Type Spacer Required Required Description in Required Appendix NUHOMS-24PT4 Yes Yes No A.1.4.1 S-100 Yes Yes No NUHOMS-32PT S-125 Yes Yes No A.1.4.2 L-100 Yes Yes No L-125 Yes Yes No -S Yes Yes No NUHOMS-24PTH -L Yes Yes No A.1.4.3 -S-LC Yes Yes No NUHOMS-32PTH Yes No No Type 1 Yes No No A.1.4.4 -S Yes No No NUHOMS-32PTH1 -M Yes No No A.1.4.5 -L No No No NUHOMS-37PTH -S Yes No No A.1.4.6 -M Yes No No NUHOMS-61BT Yes Yes No A.1.4.7 NUHOMS-61BTH Type 1 Type2 Yes Yes No A.1.4.8 NUHOMS-69BTH Yes No YesC1l A.1.4.9 RWC Yes Yes No A.1.4.9A (1) For Heat Loads Greater than 26kW NUH09.0101 A.1-9 All Indicated Changes are discussed in Enclosure 3, Item 1

MP197 Transportation Packaging Safety Analysis Report Appendix A.l.4.9A Radioactive Waste Canister TABLE OF CONTENTS Rev. IBA, 02/18 I A.l.4.9A.1 Radioactive Waste Canister Description................................................... A.1.4.9A-1 A.1.4.9A.2 RWC Contents........................................................................................... A.1.4.9A-:2 A.1.4.9A.3 References.................................................................................................. A.1.4.9A-4 LIST OF TABLES Table A.1.4.9A-1 Nominal Dimensions of the RWC................................................. A.1.4.9A-5 Table A.1.4.9A-2 Nominal Dimensions of the RWC Inner Liner.............................. A.1.4.9A-5 NUH09.0101 A.l.4.9A-i

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Appendix A.1.4.9A Radioactive Waste Canister NOTE: References in this appendix are shown as [1], [2], etc., and refer to the reference list in Section A.l.4.9A.3. A.1.4.9A.1 Radioactive Waste Canister Description The radioactive waste canister (RWC) is designed to contain dry irradiated and/or contaminated non-fuel-bearing solid materials (described further in paragraph A.1.4.9A.2). Under normal transport conditions, the canister rests on four canister rails, attached to the inside surface of an inner sleeve of the NUHOMS-MP197HB transport cask. The RWC is designed to transport its payload dry and in an air or inert gas environment. When a wet-load procedure (i.e., in-pool) is followed for cask loading, the RWC and transport cask cavities are drained and dried in order to ensure that free liquids do not remain in the package during transport. The heat generated by the contents of the RWC is transferred through the transport cask to the environment by conduction, convection and radiation. No forced cooling is required. Each RWC assembly consists of a cylindrical shell, top shield plug, outer top cover plate, bottom shield plug, and outer bottom cover plate. As shown in Table A.1.4.9A-1, the RWC system consists of three design configurations: Welded Top Shield Plug Design (RWC-W) Bolted Top Shield Plug Design (RWC-B) Dismantling and Decommissioning Design (RWC-DD) Table A.l.4.9A-1 provides the overall dimensions for each RWC configuration. The details of each configuration are included in the drawings contained in Section A.1.4.10.11 of Appendix A.1.4.10. The RWC assembly is constructed of steel materials with welded or bolted configurations that provide for handling the contents and biological shielding. The RWC assembly provides a minimum steel thickness of 1. 75 inches in the radial direction. The RWC assembly provides a minimum steel thickness of 5. 7 5 inches below the 12,ayJoad and a minimum steel thickness of 7. 00 /nche,~.Cfbor,eJhe P.(!J!/oa,d in the Clff~l. d~!~~tions'.(E~~arthin ~pot~?>fthe _Rfi:'.S ~e,s~f~iflgfn: ah d.u~,e.. ~::th1ckness 1stacp(!ptable,prQ1Jlded it meets:th~ allowabie,.stress, limits *listeil'.m'Apperid,1xJ 2.1;},:72) -*-m Material properties are listed in Chapter A.2, Table A.2-4. All internal structural components and payloads are the same or similar alloys of stainless steel or carbon steel. These materials are not subject to chemical or galvanic interaction. No hydrogen gas generation is induced by chemical, galvanic, thermal, or radiolytic reactions. All RWC welding procedures, welders, and welding are performed in accordance with the requirements of AWS D 1.1 [l] and AWS D 1. 6 [2]. All inspections are performed in accordance withAWS DJ.I [l] andAWS Dl.6 [2]. NUH09.0101 A.l.4.9A-l All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.1.4.9A.1.1 Welded Top Shield Plug Design (RWC-W) The RWC-W shell assembly is a 1.25 inch thick steel welded configuration that provides corifinement of radioactive materials, encapsulate the contents in an air or inert atmosphere, and provide biological shielding. The RWC-W shell has redundant seal welds that join the shell and the top and bottom cover plate assemblies to seal the canister. The bottom end assembly welds are made duringfabrication of the RWC-W shell. The top end closure welds are made after content loading. Both top plug penetrations (siphon and vent ports) are sealed after the RWC-W drying and baclfzlling operations are complete. An inner liner assembly that is used with the RWC-W is a 0.50 inch thick steel welded cylinder with a bottom plate. The bottom plate is designed with drain holes on the bottom to allow liquid from the inner liner to drain into the bottom of the RWC for dewatering. Four lifting lugs are provided on the inner liner for lifting the inner liner either empty or loaded. The lugs are designed,fabricated and tested to the requirements of ANSI N14.6 {3}. The inner liner is manufactured with a keyway for alignment in the outer RWC-W canister. A.1.4.9A.1.2 Bolted Top Shield Plug Design (RWC-B) The RWC-B shell and bottom are the same as the RWC-W, except the RWC-B provides an option for a bolted top shield plug and does not utilize an inner liner. The bolted top shield plug allows for multiple loadings. Both the top shield plug and outer lid are seal welded after final loading. The RWC-B cylindrical shell is 1. 75 inch thick to provide the same shielding as the RWC-W 1.25 inch thick shell used with the 0.50 inch thick inner liner. A.1.4.9A.l.3 Dismantling and Decommissioning Design (RWC-DD) The RWC-DD is a variant of the RWC design configurations that are approved for use with the MP197HB. The RWC-DD canister nominal diameter is the same as the approved RWC corifigurations, but longer in length. The RWC-DD is intended to be a reusable canister for transport of secondary liners containing waste or single use for disposal of low level waste in shallow earth disposal sites. The RWC-DD configuration@not intended for extended dry storage like the RWC -Wand RWC-B. The RWC-DD shield plug and outer lid are an integral bolted construction without seal welding an outer lid. This bolted outer lid configuration allows reuse of the canister for loading, transJ!.ort, and unloadingf9r disposal of contents resulting.ft.om des..~ioning activitie~. C!fh:,.;;;;J?J tnse#s.are alfowed:a.t the b61t zoca,tions.hr.ihe *,f(ii:~r 1ti0lJ. fihe Rfilf3'-Dlli ~ 'i1"'fuh7 YI~ '_,~ NUH09.0101 A.l.4.9A-2 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.l.4.9A.2 RWC Contents The NUHOMS-MP197HB packaging is designed to transport a payload ofup to 56.0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The safety analysis of the cask takes no credit for the containment provided by the RWC. The q14antity ofraq)oacJjye materiaUslimi{ed to a 11J(1Ximum of9Q;OOO CJ, of cobalt-:60 o~--~. equivalent, except for MP 197HB Unit OJ where the limit is reduce.to 70,.000 Ci of co_balt.:.60 or) (tquiv(llent. Equivalen{qctivitJ: Um ifs as a function ofgC1mmq pne7:gy for,~!sof(HJes. other thqn( (Co-60 are shown in Table A. 7~2cfor the 90,000 Ci limit and Table A. 7-2dfor the 70,000 Cit {limit. The quanti~of rad)oac({v~ mqterialjs lim,ited to. C!, maximum of 90,000 (i of c,qbalt-:,'?:9~ tequivalent, except/or MP 197HB Unit 01 where the limit is,;educ~ to 70, QOO C! of co~alt-§.0 or) (eHuiva/ent. Equiyqlent(!ctiv~tJJimi,t,~.as afunctfqn of gamm9 en~rf?Y Johsotopes ot~er,th,qn) ~Co-60 are shown in Table A. 7.;.2c_jor the 90;-000 Ci limit and Table A. 7-2dfor'the 70,_(!_00 Ci) (JirnHJThe radioactive material is typically in the form of neutron activated metals, or metal oxides in solid form. Surface contamination may also be present on the irradiated components. When a wet-load procedure (i.e., in-pool) is followed for cask loading, the cask cavity and RWC are drained and dried to ensure that there are no free liquids in the package during transport. The RWC shall contain dry irradiated and/or contaminated nonfuel bearing solid materials. The dry irradiated and/or contaminated non-fuel bearing solid materials'whose total RWC payload meets concentration requirements as low level radioactive waste (LLRW) per JO CFR 61.55. Waste characterization per IO CFR 61. 5 5 is the basis for demonstrating compliance with activity limits for transportation. The contents will not include liquid wastes, sludge, resins, or organic material. Waste containingfissile material is acceptable provided the quantity of fissile material is limited such that it can be exempted from being classified as fissile material per JO CFR 71.15 (e.g.,fission chambersfor in-core detectors). A.1.4.9A.2.1 Type and Form of Material The NUHOMs9-MP 197HB packaging is designed for shipment of various types of irradiated and contaminated reactor hardware. The payload will vary from shipment to shipment. Typical composition of the payload consists of the following components either individually or in combinations: The typical cobalt-60 specific activity ranges for these items are as follows:

1. BWR Control Rod Blades l.3x10 l. lxl0-2 Ci/g

2. BWR Local Power Range Monitors (LP RMs) l.Oxl 0 4.8xl 0-2 Ci/g

3. BWR Fuel Channels 7.8x10 2.0xl0-4 Ci/g

4. BWR Poison Curtains 6.2xl 0 4.0xl 0-2 Ci/g

5. PWR Burnable Poison Rod Assemblies (BPRAs) 3.8x10 l.3x10-3 Ci/g

6. BWR and PWR Reactor Vessel and Internals 2.0xl0 l.3x10-2 Ci/g A.1.4.9A.2.2 Decay Heat load The RWC heat load does not exceed 5kw, well below the limit of 26 kW limit for the DSC contents in MP 197HB without external cooling fins.

NUH09.0101 A.l.4.9A-3 All Indicated Changes are discussed in RAI 5-1 and Enclosure 3, Item 2

MPI97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.1.4.9A.2.3 Loading Components with high specific activity are generally placed near the center of the RWC. For each shipment, the RWC is normally filled to capacity, which prevents shifting of the contents during transport. If the RWC is not full, appropriate component spacers or shoring is used to prevent significant movement of the contents. NUH09.0I01 A.1.4.9A-3a

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.l.4.9A.2.4 Maximum Quantity of Material per Package (f_or96t1taj_nm~Y1t,Jthe quantity of radioactive material is limited to a maximum of 8,182 A2. The radioactive material is primarily in the form of neutron activated metals, or metal oxides in solid form. Surface contamination may also be present on the irradiated components. When a wet load procedure (i.e., in-pool) is followed for cask loading, the cask cavity and RWC are drained and dried to ensure that there are no free liquids in the package during transport. The NUHOM!fID-MP l 97HB packaging is designed to transport a payload of up to 56. 0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The payload of up to 56.0 tons includes the weight of the cask sleeve plus the RWC and its contents. (!!or shj_<!_[t!:1.,z_&)the maximlf:!!: quqpti!JL9L11<!!J:f_l!.<!L bearing radioactive material loaded into a p_gckage shall not ~XC(le<!J90}'000 Ct'of cobcilt-60 of,equivalent. The maxirnllm RW<; qlfbvf:q!!Ji) fE2ntent is redl!.,ced to 70,000,Ci ofCp-60 or equivalent when load~d in MP 197HB Unit 01.) A.l.4.9A.3 References

1.

American Welding Society, Dl.1-98, Structural Welding Code-Steel

2.

American Welding Society, Dl.6-99, Structural Welding Code-Stainless Steel

3.

ANSI N14.6, Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More, 1993 NUH09.0101 A.l.4.9A-4 All Indicated Changes are discussed in RAI 5-1 and Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Shell Thickness (in) Canister Length (in.) Outside Diameter (in.) Cavity Length (in.) Cavity Diameter (in.) t/ote: Table A.1.4.9A-1 Nominal Dimensions of the RWC RWC Design Parameters RWC-W RWC-B (l.75({U 1.75 186.50 186.50 67.19 67.19 167.30 167.30 64.69 63.69 Rev. 18D, 12/18 I RWC-DD 1.75 196 67.25 183.25 63.75 (1) The shell thickness for the RWC-W is 1. 7 5 inches, which includes the RWC-W inner liner thickness of 0.50 inches. Table A.1.4.9A-2 Nominal Dimensions of the RWC Inner Liner RWC-W Inner Liner Design Parameters Shell Thickness (in.) 0.50 Outside Length (in.) 166.30 Outside Diameter (in.) 63.69 Cavity Length (in.) 162.11 Cavity Diameter (in.) 62.69 NUH09.0101 A.l.4.9A-5 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Appendix A.1.4.10 Drawings of Transport Packaging and DSCs TABLE OF CONTENTS A.1.4.10.1 NUHOMsID-MP197HB DRAWINGS........................................................... A.1.4.10-7 A.1.4.10.2 NUHOMsID 24PT4 DSC DRAWINGS........................................................ A.1.4.10-23 A.1.4.10.3 NUHOMsID 32PT DSC DRAWINGS.......................................................... A.1.4.10-41 A.1.4.10.4 NUHOMsID 24PTH DSC DRAWINGS....................................................... A.1.4.10-76 A.1.4.10.5 NUHOMsID 32PTH DSC DRAWINGS..................................................... A.1.4.10-112 A.1.4.10.6 NUHOMsID 32PTHJ DSC DRAWINGS................................................... A.1.4.10-158 A.1.4.10.7 NUHOMsID 37PTH DSC DRAWINGS..................................................... A.1.4.10-193 A.1.4.10.8 NUHOMsID 6JBT DSC DRAWINGS........................................................ A.1.4.10-228 A.1.4.10.9 NUHOMsID 61BTH DSC DRAWINGS..................................................... A.1.4.10-243 A.1.4.10.10 NUHOMsID 69BTH DSC DRAWINGS..................................................... A.1.4.10-273 A. I. 4.10.11 Radioactive Waste Canister Drawing...................................................... A. I. 4.10-3 0 7 Editorial NUH09.0101 A.1.4.10-i

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Appendix A.1.4.10 NUHOMS-MP197HB SAR Drawings The following drawings for the NUHOMS-MP197HB Cask are included in Section A.1.4.10.1. Drawing Number MP197HB-71-1001 Rev 5 MP197HB-71-1002 Rev 8 MP197HB-71-1003 Rev 3 MP197HB-71-1004 Rev 7. MP197HB-71-1005 Rev 8 MP197HB-71-1006 Rev 5 MP197HB-71-1007 Rev 1 MP197HB-71-1008 Rev 3 I MP197HB-71-1009 Rev 4 MP197HB-71-1011 Rev 1 MP197HB-71-1014 Rev 3 Title NUHOMS-MPI97HB Packaging Transport Configuration (2 sheets) NUHOMS-MP197HB Packaging Parts List (2 sheets) NUHOMS -MPl 97HB Packaging General Arrangement (1 sheet) NUHOMS-MP197HB Packaging Cask Body Assembly (1 sheet) NUHOMS-MPl97HB Packaging Cask Body Details (3 sheets) NUHOMS-MPl97HB Packaging Lid Assembly and Details (I sheet) NUHOMS-MP197HB Packaging Regulatory Plate (I sheet) NUHOMS-MPl97HB Packaging Impact Limiter Assembly (I sheet) NUHOMS-MPl97HB Packaging Impact Limiter Details (I sheet) NUHOMS-MPl97HB Packaging Transport Configuration Outer Sleeve With Fins Option (I sheet) NUHOMS-MP197HB Packaging Internal Sleeve Design (I sheet) The following drawings for the NUHOMS24PT4 DSC are included in Section A.1.4.10.2. Drawing Number NUH24PT4-7l-I001 Rev 0 NUH24PT4-71-1002 Rev 0 NUH24PT4-71-1003 Rev 0 Title NUHOMS 24PT4 Transportable Canister For PWR Fuel Basket Assembly (5 sheets) NUHOMS 24PT4 Transportable Canister For PWR Fuel Main Assembly (8 sheets) NUHOMS 24PT4 Transportable Canister For PWR Fuel Failed Fuel Can ( 4 sheets) NUH09.0101 A.1.4.10-1 All Indicated Changes discussed in Enclosure 1 and Enclosure 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 The following drawings for the NUHOMS 69BTH DSC are included in Section A.1.4.10.10. Drawing Number NUH69BTH-71-1001 Rev 3 NUH69BTH-71-1002 Rev 3 NUH69BTH-71-1003 Rev 3 NUH69BTH-71-1004 Rev 6 NUH69BTH-71-1011 Rev 3 NUH69BTH-71-1012 Rev 4 NUH69BTH-71-1013 Rev 4 NUH69BTH-71-1014 Rev 2 NUH69BTH-71-1015 Rev 2 Title NUHOMS 69BTH Transportable Canister For BWR Fuel Main Assembly ( 4 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Basket - Shell Assembly (4 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Shell Assembly ( 4 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Alternate Top Closure (7 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Basket Assembly (5 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Transition Rail Assembly And Details (6 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Holddown Ring Assembly (2 sheets) NUHOMS 69BTH Transportable Canister For BWR Fuel Damaged Fuel Modification (1 sheet) NUHOMS 69BTH Transportable Canister For BWR Fuel Damaged Fuel End Caps (1 sheet) The following drawings for the Radioactive Waste Canister are included in Section A.1.4.10.11. Title Drawing Number NUHRWC-71-1001 Rev 4 NUHOMS System Radioactive Waste Canister (2 sheets) NUH09.0101 A.1.4.10-6 All Indicated Changes discussed in Enclosure 1 and Enclosure 2

MP197 Transportation Packaging Safety Analysis Report Rev. 12, 02/12 I A.1.4.10.1 NUHOMS-MP197HB DRAWINGS This section contains drawings for the NUHOMS -MP197HB. NUH09.0101 A.1.4.10-7

8 I H G F E D C 8 A 8 I 7 I 6 I 5 l 4 I 3 I Security Related Information on the Drawings in Section A.1.4.10.1, are Withheld per the Criteria of RIS 2005-31 7 I 6 I 5 T 4 I 3 I 2 I H G F E D C 8 A 2 I

MP197 Transportation Packaging Safety Analysis Report Rev.18A, 02/18 I A.1.4.10.11 Radioactive Waste Canister Drawing This section contains drawings for the Radioactive Waste Canister. NUH09.0101 A.1.4.10-307

8 I H G F E D C B A 8 I 7 I 6 I 5 j 4 I 3 I 2 I Security Related Information on the Drawings in Section A.1.4.10.11, are Withheld per the Criteria of RIS 2005-31 7 I 6 I 5 1 4 I 3 I 2 I H G F E D C B A

MP197'Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13.5-3 __.. _____________________. ___________________________________. _____________ Design :earameters ______ _________. Ni, Number of trunnions 2 av Vertical acceleration (g) 3 aL Longitudinal acceleration (g) 10 Thos Outer shell thickness (in) 2.75 Nb Number of bolts per trunnion 12 K Nut factor 0.135 Dbolt Bolt circle diameter (in) 21.00 Rweld Radius of outer shell at attachment block (in) 42.25 Thtool Thickness of lifting tool (in) 1.00 L.r Length of extremity flange (in) 0.63 Dcav Depth of inner trunnion cavity (in) 8.76 Di Diameter oftrunnion cavity - double shoulder trunnion (in) 5.00 Dis Diameter of trunnion cavity - single shoulder trunnion (in) 6.75 Dextl Minimum shoulder diameter - double shoulder trunnion (in) 19i°74) L-....:......:.. Dextls Minimum shoulder diameter - single shoulder trunnion (in) 9.75 L1 Outer shoulder length - double shoulder trunnion (in) 4.38 D.x12 Maximum shoulder diameter (in) 11.81 L2 Inner shoulder length - double shoulder trunnion (in) 4.56 L3 Shoulder length - single shoulder trunnion (in) 4.00 Dmax Maximum trunnion diameter (in) 25.00 Dflange Diameter of trunnion flange (in) 17.00 Thflange Thickness of flange at closure bolt circle (in) 2.68 Hmax Maximum height of attachment block (in) 7.35 Thw_ext Thickness of external weld of attachment block (in) 1.50 NUH09.0101 A.2.13.5-20 All Indicated Changes are discussed in Enclosure 2, Number 11

l\\.1Pl97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13.5-4 ..... -- *-*-------** *-... ----****-*-**----*----- ------ --------------***--- ----Dou.ble-Shoulder-Tr-u.n.n.ion-St~ess-Calculation _____ ---- ------------------------***--**-*--------- ___... Section A-A B-B C-C Jr( 2

2) n(

2

2) 7r 2

SBB =- Dext2 -Di = Sec = 4 x Dext2 = Stress _!!'4 ~4 De::_11 -Di = l] 4 area (in2)

(9.74 2 -5.0 2

) = 54.8711 7r (11.812 -5.02 )= 89.91 7r X 11.812 = 109.54 4 4 Jr( 4

4) n( 4

4) 7r 4

Moment JAA = 64 Dextl -Di = JBB =- Dext2 -Di = Ice =-xD 2 = ofinertia 64 64 ext (in4) I~ (9:744-s.oo')-44 ~(11.814 -5.04 )= 924.2 ~xll.814 =954.9 64 64 L1+ Li-(Ler+-L1+Li-Dcav)- Bending Lt -Thtool Thtool L1+L2-Thtool distance =4.38-1 = Dcav-Ler Thtool = 4.38 + 4.56 - 1 (in) =3.38=LAA = 8.76- 0.63 -1 =7.94=Lee =7.13 =LBB Bending MAA =Fv x LAA MBB=Fv x LBB Mee=Fvx Lee moment = 478,500 X 3.38 = 478,500 X 7.13 = 478,500 X 7.94 (in.lb) = 1,617,330 = 3,411,705 =3,799,290 rv = 478,500 1 Shear Fv 478,500 Fv 478,500 = = stress (ksi) lj SAA - 54.8:J SBB 89.81 Sec 109.54 I -8.7 =5.3 =4.4 = ~ MAA Dextl --x-- MBB Dext2 --x-- Mee Dex,2 --x-- Bending ]AA 2 /BB 2 Ice 2 stress (ksi) i-J,6T7,330 X 9.741 3,411,705 11.81 3,799,290 11.81 = x-- = x-- I 411.1 2 1 924.5 2 954.9 2 L_ =19.2 =21.8 =23.5. Max. Q192' +4xs.s' If stress .J21.82 +4x5.32 .J23.52 +4x4.42 intensity [ =25.9==11 =24.2 =25.1 (ksi) _.==;.___ NUH09.0101 A.2.13.5-22 All Indicated Changes are discussed in Enclosure 2, Number 11

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13.5-8

• • --* *-. - **-. -.. --* -. -- *--. -. - *--- ---. Summary* ofLifti11g-Sfresses-Dou6Ie-Sli.oiililerTrunnioiis(Jisif-- ----- -- *-*- ----** --

Calculated Allowable Margin Calculated Allowable Margin (3 e:) (Sv) (5e:) (Su) Stress intensity in section A-A 25.9 0.10 43.2 0.78 Stress intensity in section B-B 24.2 28.5 0.18 40.3 77.0 0.91 Stress intensity in section C-C 25.1 0.13 41.8 0.84 Bolt tensile stress 73.4 140.3 0.91 110.4 165.0 0.49 Stress intensity in trunnion flange 10.2 28.5 1.79 17.0 77.0 3.53 Weld bending stress 6.2 35.4 4.70 10.3 70.0 5.77 Outer cask shell stress 17.2 35.4 1.06 28.7 70.0 1.44 NUH09.0I01 A.2.13.5-26 All Indicated Changes are discussed in Enclosure 2, Number 11

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 14.75 .63

8. 7 6 -------'

¢25.00 I ¢ 17.00 _J Jl268 4.38 ----- 4.56 Figure A.2.13.5-1 Double Shoulder Trunnion Section NUH09.0101 A.2.13.5-29 All Indicated Changes are discussed in Enclosure 2, Number 11

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Appendix A.2.13.7

---

--- -- - - ----------- -- - - --- ------- ----- -MF-1-9.'.7HB-DSC-(Shell-Assembly)-Str-ucturaLEv.aluation -. -- -- --- --- ----- --- --.. --* --. - ---* -- -

TABLE OF CONTENTS A.2.13. 7.1 Introduction.................................................................................................. A.2.13. 7-1 A.2.13. 7.2 Dynamic Load Factors and Maximum Drop Accelerations........................ A.2.13. 7-4 A.2.13. 7.3 Canister Structural Analysis........................................................................ A.2.13. 7-6 A.2.13. 7.4 Stress Analysis Results............................................................................... A.2.13. 7-14 A.2.13.7.4.1 Group 1 DSC Stress Analysis Results................................................... A.2.13.7-14 A.2.13. 7.4.2 Group 2 DSC Stress Analysis Results................................................... A.2.13. 7-22 A.2.13.7.4.3 Group 3 DSC Stress Analysis Results................................................... A.2.13. 7-27 A.2.13. 7.4.4 Group 4 DSC Stress Analysis Results................................................... A.2.13. 7-32 A.2.13. 7.4.5 Limit Analysis....................................................................................... A.2.13. 7-40 A.2.13.7.4.6 Group 5RWC Stress Analysis Results.................................................. A.2.13. 7-43 A.2.13. 7.5 References.................................................................................................. A.2.13. 7-44 Table A.2.13. 7-1 Table A.2.13. 7-2 Table A.2.13. 7-3 Table A.2.13. 7-4 Table A.2.13. 7-4a LIST OF TABLES Material Properties for Steels SA-240 304, SA-479 304, SA-182 F304, & SA-336 F304............................................................. A.2.13. 7-45 Material Properties for Steel A36............................. ~.................... A.2.13. 7-46 Material Properties for B-29 Lead................................................ A.2.13. 7-47 Allowable Stress Values for Stainless Steel and Carbon Steel at 500 °F for DSCs......................................................................... A.2.13. 7-48 Allowable Stress Values for Stainless Steel at 150 °Ffor RWCs............................................................................................ A.2.13. 7-48a Table A.2.13. 7-5 Allowable Weld Stresses for DSCs................................................ A.2.13. 7-49 Table A.2.13. 7-5a Allowable Weld Stresses at 150°Ffor RWCs.............................. A.2.13.-7-49a NUH09.0101 A.2.13.7-i All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I LIST OF FIGURES Figure A.2.13. 7-1 DSC Groups 1, 2 and 3 Typical Top End 3D Finite Element Model............................................................................................. A.2.13. 7-50 Figure A.2.13. 7-2 DSC Groups 1, 2 and 3 Typical Bottom End 3D Finite Element Model............................................................................... A.2.13. 7-51 Figure A.2.13. 7-3 DSC Groups 1, 2 and 3 Typical Top End 3D Finite Element Model Mesh.................................................................................... A.2.13. 7-52 Figure A.2.13. 7-4 DSC Groups 1, 2 and 3 Typical Bottom End 3D Finite Element Model Mesh...................................................................... A.2.13. 7-53 Figure A.2.13. 7-5 DSC Group 4 Typical Top End 3D Finite Element Model............ A.2.13. 7-54 Figure A.2.13. 7-6 DSC Group 4 Typical Bottom End 3D Finite Element Model....... A.2.13. 7-55 Figure A.2.13. 7-7 DSC Group 4 Typical Top End 3D Finite Element Model Mesh............................................................................................... A.2.13. 7-56 Figure A.2.13. 7-8 DSC Group 4 Typical Bottom End 3D Finite Element Model Mesh............................................................................................... A.2.13. 7-57 Figure A.2.13. 7-8a Group 5 RWCs Typical 3D Finite Element Model Mesh............. A.2.13. 7-57a Figure A.2.13. 7-9 Group 2 DSCs -1 Load Case Deflection Plot at 75g..................... A.2.13. 7-58 Figure A.2.13. 7-10 Group 2 DSCs -2 Load Case Deflection Plot at 75g..................... A.2.13. 7-59 Figure A.2.13. 7-11 Group 4 DSCs -1 Load Case Deflection Plot at 75g......... :........... A.2.l 3. 7-60 Figure A.2.13. 7-12 Group 4 DSCs -2 Load Case Deflection Plot at 75g..................... A.2.l 3. 7-61 Figure A.2.13. 7-13 Group 2 DSCs -1 Load Case Equivalent Plastic Strain at 25g..... A.2.13. 7-62 Figure A.2.13. 7-14 Group 2 DSCs -2 Load Case Equivalent Plastic Strain at 25g..... A.2.13. 7-57 Figure A.2.13. 7-15 Group 4 DSCs-1 Load Case Equivalent Plastic Strain at 25g..... A.2.13. 7-58 Figure A.2.13. 7-16 Group 4_DSCs-2 Load Case Equivalent Plastic Strain at 25g..... A.2.13.7-59 NUH09.0101 A.2.13.7-ii All Indicated Changes are discussed in RAI 2-1

rvIP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Appendix A.2.13.7

--** ---- - ----.. --- --- -------- --ME-1-91.lIB-DSC-{ShelLAssembly)-Structural Evaluation.----- - --*- - ----*-* -- -------- -----------*-* A.2.13. 7.1 Introduction The MP J 97HB DSC and RWC shell assemblies consist of a cylindrical shell, bottom and top cover plates (inner and outer, as applicable) and bottom and top shield plugs. Each DSC shell assembly functions to support a basket assembly and confine associated fuel assemblies that are contained within the DSC shell assembly. The RWC shell assembly coefines non-fuel bearing solid materials and provides biological shielding. Multiple DSC shell assembly designs are evaluated. Each design is categorized into one of five groups based on similarity of geometry, plate thicknesses ai:id compartment payload. The five groups and the corresponding canisters are as follows: Group Grouped Canisters 1 69BTH, 37PTH, 32PTH, 32PTH Type 1, 32PTH1 2 61BT, 61BTH Type 1 3 61BTH Type 2, 32PT, 24PTH 4 24PT4, 24PTH-S-LC 5 RWC-W, RWC-B, RWC-DD For each group, the bounding payload weight (basket plus fuel assembly) and bounding design configuration are used for the analyses. DSC and RWC design features and dimensions are provided in Appendices A.1.4.1 through A.1.4.10. Appendices A.1.4.1 through A.1.4.9 provide detailed descriptions of each DSC. Appendix A. I. 4. 9A provides a detailed description for the RWC. Appendix A.1.4.10 contains reference drawings for all of the above listed DSCs and RWCs. The following paragraphs highlight the DSC and RWC similarities and the bases for the five canister groups. Group 1 The top and bottom end assembly dimensions for the Group 1 DSCs are given in the following table. Group 1 DSC Top and Bottom End Assembly Dimensions (in.) 32PTH Component 32PTH Typel 32PTH1-S 32PTH1-M 32PTH1-L 69BTH 37PTH Outer Top Cover 2.00 2.00 2.00 2.00 2.00 2.00 2.00 Inner Top Cover 2.00 2.00 2.00 2.00 2.00 2.00 Top Shield Plug 10.00 8.00 8.00 8.00 6.00 5.75 5.75 Inner Bottom Cover 1.75 2.25 2.25 2.25 2.25 1.75 1.75 Bottom Shield Plug 5.25 4.50 4.50 4.50 2.25 3.50 3.50 Outer Bottom Cover 1.75 2.00 2.00 2.00 2.00 2.00 2.00 NUH09.0101 A.2.13.7-1 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Group 5 -*----*------*---The* top and 5ottom endassem5ly dimeriszorisf6rthe*oroup-S-RWCs7ire-g1ven-iirtnefollowzng-*-- - ----*-- --- table. Group 5 RWC Top and Bottom End Assembly Dimensions (in.) Component RWC-W RWC-B RWC-DD Outer Top Cover Plate 2.00 2.00 2.00 Top Shield Plug 5.00 5.00 5.00 Bottom Shield Plug 3.75 3.75 3.75 Outer Bottom Cover Plate 2.00 2.00 2.00 The RWC is analyzed using a three-dimensional (3D) 180° half-symmetric finite element. The bottom assembly of the 3D finite element model conservatively includes only the outer bottom cover plate. A single enveloping finite element (FE) model encompassing all three design options is modeled for all structural analyses. NUH09.0101 A.2.13. 7-3a All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.2.13.7.2 Dynamic Load Factors and Maximum Drop Accelerations A Dynamic Load Factors Two load cases are considered for development of dynamic load factors and maximum drop accelerations; one due to longitudinal loading (end drop) and one due to transverse loading (side drop). During an end drop, the fundamental natural periods of the DSC and RWC components are taken to be that of simply supported cylindrical shells without axial constraint, under longitudinal vibration. During a side drop, the fundamental natural period of the canister shell is taken to be that of a cylinder in an ovalling mode and a simply supported cylindrical shell without axial constraint. Since the canister is not modeled in detail in the transport cask dynamic analysis, it is necessary to transfer the loads from the cask dynamic analysis model to the detailed models of the canister. The canisters are evaluated using quasi-static analyses with a dynamic load factor (DLF) computed from the transient dynamic analysis. NUH09.0I01 A.2.13.7-4 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report I Rev. 18D, f2/18 The development of the dynamic load factor applicable to each canister type is described in Appendix A.2.13.9. The results are summarized in the following table. Canister Dynamic Load Factor Results Summary for Each Canister MP197-HB I 61BT/61BTH/ 24PTHF/ 32PTH/ I Transport Package 61BTHF 69BTH 24PTH 24PT4 32PT 32PTH1 37PTH RWC-W RWC-B RWC-DJ? Drop Orientation i DLF DLF DLF DLF DLF DLF DLF DLF DLF DLF I Normal Condition 1.32 1.31 1.32 1.18 1.33 1.31 1.30 1.20 1.15 1.12 i Canister End Drop Accident Condition 1.17 1.16 1.16 1.17 1.17 1.16 1.15 1.16 1.21 1.22 I Canister End Drop I I I I Normal Condition I Canister Side Drop 1.27 1.29 1.30 1.26 1.30 1.26 1.31 1.27 1.18 1.26 i Accident Condition I Canister Side Drop 1.01 1.00 1.00 1.01 1.00 1.01 1.02 1.01 1.01 1.01 i The bounding dynamic load factor corresponding to each group of canister types is used to bound the analysis of the canister in e~ch group. The bounding DLFs used in each group are summarized in the table below. Bounding Dynamic Load Factor for Each Group DLF DLF DLF DLF DLF MP197-HB Transport Package Drop Orientation Group 1 Group2 Gr011p 3 Group4 Groups DSCs DSCs DSCs DSCs RWCs Normal Condition Canister End Drop 1.31 1.32 1.33 1.32 1.20 Accident Condition Canister End Drop 1.16 1.17 1.17 1.17 1.22 Normal Condition Canister Side Drop 1.31 1.27 1.30 1.30 1.27 Accident Condition Canister Side Drop 1.02 1.01 1.01 1.01 1.01 NUH09.0101 A.2.13. 7-4a All Indicated Changes are discussed in RAI 2-1

l\\1Pl97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I B. Development of Maximum Accelerations --- ---- --- -*This*-section calculatestherequtre-a-accelerations-for eachoflne--r>SCana'RWC-groups and-for- ----1---- --- use in the canister structural evaluations. The baseline g loads are calculated from LS-DYNA analyses of the transport cask and are described in Appendix A.2.13.12, Section A.2.13.12.10. The results are as follows: Normal Condition Canister End Drop: 18g Accident Condition Canister End Drop: 55g Normal Condition Canister Side Drop: 19g Accident Condition Canister Side Drop: 55g Multiplying the above accelerations by the bounding DLFs in each DSC and RWC group results in the required accelerations to be used for the DSC and RWC analyses. Actual acc,eleration values used in the analyses are greater than the calculated values. The calculated required acceleration values and the values used in the analyses are shown in the following table. Using 75g for DSC and RWC canister HAC end drop and side drop analyses bound all other drop orientations. Bounding g Loads for Each Group MP197-HB Transport Package Group 1 Group2 Group3 Group 4 Groups Acceleration Calculated Required Nonna! Condition 18xl.31= 18xl.32= 18xl.33= 18xl.32= lBxl.20= Canister End Drop Acceleration 23.58 23.76 23.94 23.76 21.60 Baseline g load for 30 30 30 30 25 Canister NCT End Drop Acceleration Calculated Required Accident Condition 55xl.16= 55xl.17= 55xl.17= 55xl.17= 55xl.22= Canister End Drop Acceleration 63.8 64.35 64.35 64.35 67.10 Baseline g load for 75 75 75 75 75 Canister RAC End Drop Acceleration Calculated Required Nonna! Condition 19xl.31= I9xl.27= I9xl.30= I9xl.30= 19xl.27= Canister Side Drop Acceleration 24.89 24.13 24.70 24.70 24.13 Baseline g load for 30 25 25 25 25 Canister NCT Side Drop Acceleration Calculated Required Accident Condition 55xl.02= 55xl.01= 55xl.Ol= 55xl.Ol= 55xl.01= Canister Side Drop Acceleration 56.l 55.55 55.55 55.55 55.55 Baseline g load for 75 75 75 75 75 Canister RAC Side Drop Acceleration NUH09.0101 A.2.13.7-5 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12118

• 1 A.2.13.7.3 Canister Structural Analysis Finite element analyses are performed in order to quantify stresses in the DSCs generated by transport loads. The applied loads considered are normal and accident condition top end, bottom end, and side drops, combined with internal and external pressures and temperature distributions (thermal expansion stresses). For the DSC, several finite element models are used to evaluate stresses for the normal and accident loads: 180 degree 3D models are used for side drop analyses; 2D axisymmetric models are used for end drop and thermal expansion analyses.

Elastic material properties are used for normal condition stress analyses. Elastic-plastic material properties are used for normal condition limit load analyses and accident condition stress analyses. A single 180° 3D model that envelops the three RWC designs is used for the bottom end and side drops. The top end drops are bounded by the bottom end drops as the top assembly is heavier than the bottom assembly. Elastic and elastic-plastic material properties are used for RWC normal and accident condition stress analyses respectively. A Material Properties Steel Material Properties for Group 1 through 4 DSCs Material properties and allowable stresses for normal (NCT) and accident (RAC) drop analyses are based on 500 °F which bound-40 °F, -20 °F, and 100 °F ambient conditions. Thermal expansion analyses are based on temperature-dependent material properties shown in Tables A.2.13.7-1 and A.2.13.7-2 [3]. For the accident condition side drop cases where elastic-plastic analyses are performed, the tangent modulus is taken as 5% of the elastic modulus. Lead Material Properties for Group 4 DSCs The Group 4 DSCs have lead shield plugs instead of steel shield plugs. Material properties for the normal and accident drop analyses are based on 500 °F which bounds the maximum DSC temperatures. For accident condition load cases, dynamic stress-strain properties are used for lead. Material properties oflead are shown in Table A.2.13.7-3 [6], [7]. Steel Material Properties for Group 5 RWCs Material properties and allowable stresses for normal (NCT) and accident (HAC) drop analyses are based on 150 °F. For the accident *condition side and end drop cases where elastic-plastic analyses are performed, the tangent modulus is taken as 5% of the elastic modulus. Temperature-dependent material properties are shown in Table A.2.13. 7-1. NUH09.0101 A.2.13.7-6 All Indicated Changes are discussed in RAI 2-1

l\\1P197 Transportation Packaging Safety Analysis Report Rev. 18D, 12118 I B. Design Criteria --** ** *** - -** - -****- -The DSCs *steel component stresses are compared w1ththe**anowable stresses set forth byAsME* 1-- -* **-- ****** - *-** B&PV Code Subsection NB [1]. The allowable stress values at 500 °P for the steel components are summarized in Table A.2.13.7-4. Closure weld stress allowables are based on ISG-15 [5], which requires a design stress reduction factor to account for weld imperfection or flaws and recommends a stress reduction factor of 0.8 based on multi-level PT examination. The corresponding values at 500 °P are summarized in Table A.2.13.7-5. If the allowable stress limits are exceeded, a simplified fatigue analysis per NB-3228.5 [1] is performed. The RWCs steel component stresses are compared with the allowable stresses set forth by ASME B&PV Code, Section III, Subsection NF [9] for normal (Level A) and ASME B&PV Code, Section III, Appendix F of [2] for accident (Level D) conditions. The allowable stress values at 150 °F for the steel components and welds are summarized in Table A.2.13.7-4A and Table A.2.13.7-5A, respectively. C. Loading Conditions The load cases considered for DSCs are normal and hypothetical accident condition drops, pressure loads, and temperature distributions (thermal expansion stresses). The normal condition drop loads are combined with internal and external pressure and the 100 °P and -20 °P ambient environment thermal loads. The accident condition drop loads are combined with internal and external pressure. For the RWCs, the load oases considered are normal and hypothetical accident condition drops. Since the contents of the RWC include only non-fuel-bearing solid materials, internal pressure and thermal loads are not considered. The following tables summarize both normal and accident condition DSC and RWC individual load cases. NUH09.0101 A.2.13.7-6a All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 DSC and RWC Normal Condition (NCT) Load Cases

• • - -Service --

Loadin!!

• Analysis Tvoe Level Load Analvsis Method Hot Environment Elastic Analysis A

Hot Ambient Finite Element Analysis Thermal Load (2D, axisymmetric model) Cold Environment Elastic Analysis A Cold Ambient Finite Element Analysis Thermal Load (2D, axisymmetric model) Internal Pressure Elastic Analysis A Internal Pressure Finite Element Analysis (included in drop analyses) External Pressure Elastic Analysis A External Pressure Finite Element Analysis (included in drop analyses) 1 Foot Elastic Analysis A Lateral g-Load Finite Element Analysis Side Drop (3D, 180 deg. model) 1 Foot Top Elastic Analysis A Axial g-Load Finite Element Analysis End Drop (2D, axisymmetric model) 1 Foot Bottom Elastic Analysis A Axial g-Load Finite Element Analysis End Drop (2D, axisvmmetric model) Note: RWC is evaluated for I Foot Side Drop and Bottom End Drop only. DSC and RWC Accident Condition (HAC) Load Cases Service Loading Analysis Type Level Load Analysis Method 30 Foot Elastic-Plastic D Lateral g-Load Finite Element Analysis Side Drop Analysis (3D, 180 deg. model) 30FootTop Elastic Analysis D Axial g-Load Finite Element Analysis End Drop (2D, axisymmetric model) 30 Foot Bottom Elastic Analysis D Axial g-Load Finite Element Analysis End Drop (2D, axisvmmetric model) Note: RWC is evaluated for 30 Foot Side Drop and Bottom End Drop only. The individual loads are combined as shown in the following tables, DSC and RWC Normal Condition (NCT) Load Combinations Load Case 1 2 3 4 5 6

• Notes:

1.

2.

3.

Individual Loads 2sg<1> 30g 30g(3) 15 psig<2> 15 psig Side Top End Bottom Internal External Hot Ambient Drop Drop End Drop Pressure Pressure Environment X X X X X X X X X X X X X X X The Group 1 DSC analyses conservatively used 30g for the normal side drop. The internal pressures used in the analyses are as follow: Group 1 DSCs: 30 psig Group 2 DSCs: 15 psig Group 3 DSCs: 15 psig Group 4 DSCs: 20 psig Cold Ambient Environment X X X Group 5 RWCs: No internal or external pressure and hot or cold ambient environment. No top end drop is required as bottom end drop is a bounding scenario. Bottom end drop for RWC is pe,formedfor 25g loads. NUH09.0101 A.2.13.7-7 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I DSC and RWC Accident Condition (HAC) Load Combinations - -- -- *- -*-- - -- --- - ---- --- ------ -- ----- ------- - Individual-Loads-- --- ---- --------- ---- --- - ---- ------ 75g 75g 75g 15 psig 15 psig Load Side Top End Bottom Internal External Case Drop Drop End Drop Pressure Pressure 7 X X 8 X X 9 X X 10 X X 11 X X 12 X X Note:

1.

The internal pressures used in the analyses are as follow: Group 1 DSCs: 30 psig Group 2 DSCs: 15 psig Group 3 DSCs: 15 psig Group 4 DSCs: 20 psig Hot Ambient Cold Ambient Environment Environment X X X X X X Group 5 RWCs: No internal pressure *and hot or cold ambient environment. No top end drop is required as bottom end drop is a bounding scenario. D. Finite Element Analysis Finite Element Model Finite element models for the DSCs are constructed to evaluate stresses for the normal and accident loads using ANSYS computer program [4]. A separate set of models is used for side drop analyses of the top end and bottom end of each DSC group. For side drop load combinations, 180 degree 3D models are used. For end drop load combinations and thermal expansion stresses, 2D axisymmetric models are used. The models for end drop loading are extended to include the full length of the DSC. For the RWC group, a 3D 180° half-symmetric finite element model encompassing all three design is used. The DSC 3D finite element models are developed using SOLID45 solid elements and 3D point-to-point CONTAI 78 and CONTAC52 contact elements. Contact between cover plates and shield plugs are modeled using CONTAl 78 elements. The RWC finite element model is developed usingS0LIDJ85 solid elements, surface-to-surface CONTA173 and TARGE170 elements, and 3D node-to-node CONTAJ 78 elements. The initial gaps between these components are considered to be closed. In addition to contact elements at the interface of the cylindrical shell and shield plates, or between plates, the models include CONTAl 78 contact elements at the interface of the cylindrical shell to the cask. The nodes of these elements are located at the OD of the cylindrical shell and ID of the cask. The initial locations of these nodes define the nominal centered gap between cask and canister. The gaps at cask rails, and between the canister and the cask, are input by contact element real constants. The contact element nodes located at the ID of the cask are held fixed in all directions, simulating a rigid cask on which the canister drops. NUH09.0101 A.2.13.7-8 All Indicated Changes are discussed in RAI 2-1 I I

Iv.lP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I

• For DSC end drop load cases and thermal expansion stresses, 2D axisymmetric finite element

... _ *-*-...... _.m.Qde.ls_ar_e_dev_elop.e_d_u_sing_ANSYS..PLANE42_elements.. __ Contac_t_hetw.e_en_cml'..erplat.es_and__ ___ _ ______. shield plugs are modeled using CONTAC12 elements. For the RWC bottom end drop a 3D finite element model is developed The interface between the outer bottom cover plate and the cask is modeled with 3D node-to-node contact elements (CONTAJ 78) acting in the axial direction. The welds/or the DSCs between the shell and inner cover plates are modeled by coupling the. contacting nodes in all directions. Welds between the shell and the outer top cover plate and between shell and outer bottom cover are modeled with PLAN42 elements. For the RWC, the weld between the shell and the outer top cover plate and between the shell and the outer bottom cover plate are modeled by merging the nodes. Symmetry boundary conditions are defined for all nodes at symmetry planes. NUH09.0101 A.2.13.7-Ba All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/181 Geometry plots of the DSCs finite* element analytical models are given in Figures A.2.13. 7-1 __. __________ *-- _ through A.2._.11. 7:--8.._ fJe.ome_tzy_plot;_of_tbe_RWCs_fi1J..ite __ e_le.ment analy.tic_aLmode_l_ is_pr.o_vided frL ________.. ______ _ Figure A.2.13.7-8A. Load Cases Accelerations used in the analyses are provided in Section A.2.13.7.3. In general, accelerations of 25g, 30g and 75g are defined for the normal side drop, normal end drop, and accident drops (side and end), respectively, in the appropriate direction for each of the drop conditions. The exception to this is for the Group 1 DSCs, where a normal side drop acceleration of 30g was conservatively used (see Section A.2.13.7.3) and for Group 5 RWCs, where a normal bottom end drop acceleration of25gwas used. Load cases used in the analyses are shown in the following tables. Load Cases for Side Drop Normal Condition of Transport (NCT) Load Case Service Number.. Loadine: Condition Level Case Description lNCT Top End DSC Model Lateral Load DSC in Cask, horizontal, supported on side, + Internal Pressure A (1) (Top Notails IP) Impact away from transport cask rails. 2NCT Top End DSC Model Lateral.Load DSC in Cask, horizontal, supported on side, + External Pressure A (2) (Top Norails EP) Impact away from transport cask rails. 3NCT Top End DSC Model Lateral Load DSC in Cask, horizontal, supported on side, + Internal Pressure A (1) (Top Rails IP) Impact onto the two transport cask rails. 4NCT Top End DSC Model Lateral Load DSC in Cask, horizontal, supported on side, + External Pressure A (2) (Top Rails EP) Impact onto the two transport cask rails. 5NCT Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + Internal Pressure A (1) (Bottom Norails IP) Impact away from transport cask rails. 6NCT Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + External Pressure A (2) ffiottom Norails EP) Impact away from transport cask rails. 7NCT Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + Internal Pressure A (1) (Bottom Rails IP) Impact onto the two transport cask rails. 8NCT Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + External Pressure A (2) (Bottom Rails EP) Impact onto the two transport cask rails. 9NCT RWC Model with Lateral Load RWC in Cask, horizontal, supported on and No Internal or External A {I) or (2) Pressure side, Impact onto two transpo,:t cask rails. IONCT RWC Model with Lateral Load RWC in Cask, horizontal, supported on {I) or (2) and No Internal or External A side, Impact away from transport cask Pressure rails. ~~

• • Number in ( ) represents the DSC normal condition load combination number shown in previous table. For the RWC, the load combination (]) or (2) are the same as there is no pressure and hot or cold ambient environment.

NUH09.0I01 A.2.13.7-9 All Indicated Changes discussed in to RAI 2-1

l'vfP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12118 I Load Cases for Side Drop Hypothetical Accident Condition (RAC) --Load Gase- --- -- ---- -- ----Service- -. ---- - ----- -- - --- Number** Loadin2 Condition Level Case Description IHAC Top End DSC Model Lateral Load+ DSC in Cask, horizontal, supported on side, Internal Pressure D (7) (Top Norails IP) Impact away from transport cask rails. 2HAC Top End DSC Model Lateral Load+ DSC in Cask, horizontal, supported on side, External Pressure D (8) (Top Norails EP) Impact away from transport cask rails. 3HAC Top End DSC Model Lateral Load + DSC in Cask, horizontal, supported on side, Internal Pressure D (7) (Top Rails IP) Impact onto the two transport cask rails. 4HAC Top End DSC Model Lateral Load+ DSC in Cask, horizontal, supported on side, External Pressure D (8) (Top Rails EP) Impact onto the two transport cask rails. 5HAC Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + Internal Pressure D (7) (Bottom Norails IP) Impact away from transport cask rails. 6HAC Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + External Pressure D (8) (Bottom Norails EP) Impact away from transport cask rails. 7HAC Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + Internal Pressure D (7) (Bottom Rails IP) Impact onto the two transport cask rails. 8HAC Bottom End DSC Model Lateral DSC in Cask, horizontal, supported on side, Load + External Pressure D (8) (Bottom Rails EP) Impact onto the two transport cask rails. 9HAC RWC Model with Lateral Load and D RWC in Cask, horizontal, supported on side, (7) or (8) No Internal or External Pressure Impact onto two transport cask rails. JOHAC RWC Model with Lateral Load and D RWC in Cask, horizontal, supported on side, (7) or (8) No Internal or External Pressure Impact awav from transoort cask rails. Note:

• • Number in ( ) represents the DSC accident condition load combination number shown in previous table. For the RWC, the load combination (7) or (8) are the same as there is no pressure and hot or cold ambient environment.

Load Cases for End Drop Normal Condition of Transport (NCT) Load Case Service Number.. Loadin2 Condition Level Case Description lNCT(ED) Lid End Drop, Axial Load + Internal Cask vertical, supported at top, axial Pressure A (3) (Top End Drop IP) acceleration + internal pressure. 2NCT(ED) Lid End Drop, Axial Load+ Cask vertical, supported at top, axial External Pressure A (4) (Top End Drop EP) acceleration + external pressure. 3NCT(ED) Bottom End Drop, Axial Load+ Cask vertical, supported at bottom, axial Internal Pressure A (5) (Bottom End Drop IP) acceleration + internal pressure. 4NCT(ED) Bottom End Drop, Axial Load+ Cask vertical, supported at bottom, axial External Pressure A (6) (Bottom End Drop EP) acceleration + external pressure. 5NCT(ED) Bottom End Drop, Axial Load A Cask vertical, supported at bottom, axial (5) or (6) acceleration. Note:

• • Number in ( ) represents the DSC normal condition load combination number shown in previous table.

For the RWC, the load combination (5) or (6) are the same as there is no pressure and hot or cold ambient environment. NUH09.0I01 A.2.13.7-10 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 Load Cases for End Drop Hypothetical Accident Condition (HAC) - l,oad-ease-* - -- --*- ---- **---------- --- *-- - -*- *-Service- - -- ---- ---- ---- -*-- *------* *---*- - ------- -- Number.. Loadine Condition Level Case Descriotion lHAC(ED) Lid End Drop, Axial Load + Cask vertical, supported at top, axial Internal Pressure D (9) (Top End Drop IP) acceleration + internal pressure. 2HAC(ED) Lid End Drop, Axial Load + Cask vertical, supported at top, axial (10) External Pressure D acceleration + external pressure (Too End Drop EP) 3HAC(ED) Bottom End Drop, Axial Cask vertical, supported at bottom, axial Load + Internal Pressure D (11) (Bottom End Drop IP) acceleration + internal pressure. 4HAC(ED) Bottom End Drop, Axial Cask vertical, supported at bottom, axial Load + External Pressure D (12) (Bottom End Drop EP) acceleration + external pressure. 5HAC(ED) Bottom End Drop, Axial D Cask vertical, supported at bottom, axial (11) or (12) Load acceleration. Note:

• • Number in ( ) represents the DSC accident condition load combination number shown in previous table.

For the RWC, the load combination (11) or {I 2) are the same as there is no pressure and hot or cold ambient environment. For the side drop and end drop analyses, the weight of the internals (basket+ fuel assemblies) is accounted for by applying a pressure to the inner surface of the canister. Bounding weight is used for the analysis of each group. The internal weights used for the analyses are listed in the following table. Group 1 Group2 Group3 Group4 Group 5 DSCs DSCs DSCs DSCs RWCs Bounding Calculated 85,000 lb. 67,000 lb. 77,000 lb. 68,000 lb. 71,000 lb. Internal Weight for the DSCs in the Group Internal Weight used for 90,000 lb. 80,000 lb. 80,000 lb. 90,000 lb. 71,000 lb. the Canister Analysis in Each Group E. Side Drop Load Analysis For the DSC side drop load cases away from"tl"ansport cask rails (lNCT, 2NCT, 5NCT, 6NCT, lHAC, 2HAC, 5HAC & 6HAC), inertia loads for canister internals is accounted for by applying a cosine varying pressure on the inside surface of the canister shell. For the RWC, inertia loads for canister internals are accounted for by applying a cosine varying pressure on the inside surface of the shell for side drop load cases. Assuming that the canister internals react upon a 90° arc of the inside surface, then the inertial load of the internals, Pee), which varies with angle, 8, (8 = 0 is at the impact point for the DSCs and at the symmetry plane for the RWC), is governed by the following expression: PeO) = P max cos(2B) c -45 ° < e < 45 ° ) Where Pmax is the maximum pressure at B = 0. Assuming the axial length of the applied load is L, the inside radius of the canister shell is R, and the load distribution, Pee) above, then the total inertial load in the drop direction generated by the internals, F, is the following: NUH09.0I01 A.2.13.7-11 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Therefore, P max for Hypothetical Accident Condition (HAC) is: I ( ) 1 -1 3.7l" SID - P = 6750000 4 +sin.7l" = 1285.7 si max (162.00)(34".375) 3 ( 4 J p Therefore, the equivalent pressure applied on the canister inside shell surface for load cases away from transfer cask rails is: Pee)= 514.3 cos(28) and 1285.7 cos(28), respectively, for the Group 1 DSC NCT and HAC load cases where(}= angle from the bottom (B = 0 for a half symmetric model) of the horizontal canister shell to the center of the finite element model, up to 45°. For the DSC and RWC side drop load cases onto the two transport cask rails, inertia loads for the basket assembly are accounted for by applying an equivalent pressure onto the first ( or innermost) rail only. The base value of pressure, at lg load, is obtained after doing a few iterations with only pressure load on the first rail (simulating the basket assembly weight) and then checking the reaction loads in appropriate direction. The value obtained for the Group 1 DSCs is 93.6 psi which is then multiplied by the appropriate g loads. For the RWC, inertia loads for canister internals are accounted for by applying a cosine varying pressure on the inside surface of the shell for side drop load cases onto the cask rails. F. End Drop Load Analysis The weight of the canister internals (basket and fuel assemblies) during end drop is accounted for by applying equivalent pressures on the supporting surfaces of the (DSC and RWC) canister components. For example, the weight of the canister internals used in the end drop analyses for the Group 2 DSCs is conservatively taken to be 80,000 lb. The corresponding pressure loads equivalent to the inertial load of the internals at 30g and 75g for the NCT and HAC end drops are: P= 23.2075 x 30g = 696.22 psi p = 23.2075 X 75g = 1740.56 psi [For NCT at 30g] [For HAC at 75g] For end drop buckling analyses, where g-loads exceed 75 g, the g values and corresponding canister internals loads are appropriately increased according to the formula shown above for accelerations beyond the 75 g load. G. Internal and ~xternal Pressure Analyses Internal and external pressures are applied to the appropriate surfaces of the cylindrical shell and cover plates using ANSYS pressure loading on the solid element surfaces. No internal or external pressure is applied for the RWC NCT and HAC cases. H. Temperature (Thermal Expansion) Analysis Temperature distributions, from Chapter A.3, are conservatively mapped onto the DSC structural models and stresses due to temperature distributions are calculated. There are no thermal loads for the RWC. NUH09.0101 A.2.13.7-13 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12118 I A.2.13.7.4 Stress Analysis Results - - For the DSCs, the maximum stress intensities-in-the canister are extractec:ffrom the ANSYS --------- i-- -- --- ---- - results, for the applicable load combinations. These stresses are compared to the normal and accident condition code allowable stress intensities. For the RWC, the linearized stress intensities (membrane and membrane+bending) are extracted from the ANSYS results,for the applicable load combinations. These stresses are compared to the normal and accident condition code allowable stress intensities. The following notation is applicable to the tables of this Section: Top: Bot: IP: EP: No-P: NoRails: Rails: NCT: RAC: TSA: S.No.: ED: A.2.13.7.4.1 Top End Model Bottom End Model Loading with internal pressure Loading with external pressure Loading without pressure Side drop away from cask rails Side drop on cask rails Loading with normal side drop acceleration Loading with accident side drop acceleration Bounding thermal stress analysis Section or component ID End Drop Group 1 DSC Stress Analysis Results A. NCT Side Drop Results - Group 1 DSCs A.1 Top End Model Stress Evaluation The following tables summarize the linearized bounding stress for the main DSC components for the NCT side drop load combinations stress results for the Group 1 DSCs Top End Model. NCT Loads Maximum Stress Intensities for Group 1 DSCs - Top End Model (Away From Impact Zone) Service Level A: Top End Model Pm Pm+Pb S.No. Component (17.5 ksi) (26.3 ksi) 1 Cylindrical Shell 14.4 24.5 2 Outer Top Cover Plate 4.8 9.1 3 Inner Top Cover Plate 5.7 10.2 4 Shield Plug 4.4 4.9 5 Support Ring 8.8 12.8 Note: The corresponding allowable stress limits are shown in brackets NUH09.0101 A.2.13.7-14 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Limit Load Result Summary for Group 2 and 4 DSCs Lower Load Case Bound Limit Actual Factor of Collapse LoadC2) g-Load Safety Load(l) Group 2 DSCs -1 75 (2/3)*75 = 50 25 2.00 Group 2 DSCs -2 75 (2/3)*75 = 50 25 2.00 Group 4 DSCs -1 75 (2/3)*75 = 50 25 2.00 Group 4 DSCs -2 75 (2/3)*75 = 50 25 2.00 Notes:

1.

This is a conservative estimate of the lower bound collapse load, since actual collapse was not achieved in the analysis.

2.

Per NB-3228.1, the limit load is 2/3 of the lower bound collapse load. A.2.13.7.4.6 Group 5 RWC Stress Analysis Results A. NCT Side Drop Results - Group 5 RWCs The following tables summarize the linearized bounding stress intensities for the main RWC components for the bounding NCT side drop load combination 9NCT and 1 ONCT for the Group 5RWCs. NCT Loads Maximum Linearized Stress Intensities for Group 5 RWCs Service Level A (JJ S.No Component Pm Pm+Pb Pt+Pb (18.5 ksi) (27.5 ksi) (53.4 ksi) 1 RWCShell 5.2 15.5 28.6 2 Outer To-,; Cover Plate 10.0 10.0 38.0 3 Outer Bottom Cover Plate 16.3 18.0 N/Ar2J Note:

1.

The corresponding stress limits are shown in the brackets.

2.

Allowable limits of Primary stresses (Pm+P,) will bound the allowable limits for the 'Primary+ Secondary' stresses. Stresses Results Away From the Impact Zone All the maximum membrane stress intensities and maximum membrane plus bending stress intensities are within the code allowable membrane and membrane plus bending stress intensity limits. Pm+Pb=l8.0 ksi <1.5S=27.5 ksi Therefore, the combined stress intensity meets the code allowable stresses. NUH09.0I01 A.2.13.7-43 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Stresses Results at the Impact Zone

• *All thelocal maximum membrane stress intensities and locat maximum membrane plusbendzng-stress intensities are within the code allowable local membrane and local membrane plus bending stress intensities limits for the RWC shell and outer top cover plate.

Pz+Pb=38.0 ksi < 2Sy=53.4 ksi Therefore, the combined stress intensity meets the code allowable stresses. A.I Weld Stresses due to NCT Side Drop Loads - Group 5 RWCs The RWC incorporates closure welds between the outer top cover plate and cylindrical shell and outer bottom cover plate and the cylindrical shell. The maximum weld stresses occur at the outer top cover plate; therefore, weld stresses are reported only at this location. Weld stresses are below the corresponding stress limits. Weld Stress Intensity-Cylindrical Shell & Outer Top Cover Plate (NCT) - Group 5 RWCs Service Level A: Normal Conditions of Transport (25g) Load Case Load Case Detail Top Cover Plate (ksi) Weld Stress Limit (ksi) 9NCT On Rails 21.2 24.0 lONCT Away from Rails 20.5 24.0 As shown in the above tables, all weld stresses are below the corresponding stress limits. B. HAC Side Drop Results - Group 5 RWCs The following table summarizes the bounding HAC side drop results for load combinations 9HAC and IOHACfor the Group 5 RWCs. The tables show the maximum linearized stress intensities irrespective of impact location. Stress Intensities for Side Drop HAC - Group 5 RWCs Load Case Load Case Detail Max. Stress Intensity [ksi] Allow. Allow. Pm Pm+Ph 9HAC On Rails 44.2 51.1 65.7 lOHAC Away from Rails 41.3 51.1 65.7 NUH09.0101 A.2.13. 7-43a All Indicated Changes are discussed in RAI 2-1

1\\t1Pl97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I B.1 Weld Stress due to HAC Side Drop Loads - Group 5 RWCs Thefollowmg table summarizes the bounding HAC side drop weld stress resliltsforload*- ----- -- ---- -- - combinations 9HAC and JOHACfor the Group 5 RWCs. The maximum weld stresses occur at the outer bottom cover plate; therefore, weld stresses are reported only at this location. Weld Stress - Side Drop HAC - Group 5 RWCs Service Level D: Hypothetical Accident Side Drop Condition (75g) Load Case Load Case Detail Outer Bottom Cover Plate Weld Stress Limit (ksi) (ksi) 9HAC On Rails 18.5 21.4 lOHAC Away from Rails 17.5 21.4 Note: Weld stresses are conservativery compared against the base metal allowables. As shown in the above table, weld stresses are below the corresponding stress limits. C. NCT End Drop Results - Group 5 RWCs The following table summarizes the NCT end drop stress results for load combination 5NCT(ED) for the Group 5 RWCs. The stresses reported in this table are the maximum irrespective of the impact location. Stress Intensities for End Drop NCT - Group 5 RWCs Service Level A : End Drop Normal Condition of Transport (25g) Stress Limits [ksi] Load Case Load Case Detail Max. Stress Intensity [ksi] Allow. Allow. Pm Pm+Ph 5NCT(ED) Bottom End Drop 2.1 18.5 27.5 The weld stresses for the End Drop NCT load combination are bounded by the Side Drop NCT load combinations. D. HAC End Drop Results - Group 5 RWCs The following tables summarize the bounding HAC end drop results for load combination 5HAC(ED) for the Group 5 RWCs. The tables show the maximum linearized stress intensities per component. Stress Intensities for End Drop NCT - Group 5 RWCs Service Level D : End Drop Hypothetical Accident Condition (75g) Stress Limits [ksi] Load Case Load Case Detail Max. Stress Intensity fksi] Allow. Allow. Pm Pm+Pb 5HAC(ED) Bottom End Drop 6.4 51.1 65.7 The weld stresses for the end drop HAC load combination are bounded by the side drop HAC load combinations. NUH09.0101 A.2.13.7-43b All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I E. HAC End Drop Buckling Results - Group 5 RWCs - -* * -- -- *Fvrthe-emJ-drop--JtA.Clvad ca:re, -the*R-WC:finite-elemenrmvdelinputlot1d"ts irfcrec1s1RJ-11p-w * -- - - -- --- 150g. The last converged load step is at 150g without experiencing any buckling. Two-thirds of 15 Og define the maximum compressive load which is 1 OOg as per the code criteria provided in [2}. The maximum compressive load of 1 OOg is higher than the accident design g load of 7 5 g of Group 5 RWCs. Therefore, the RWC meets the buckling criteria. NUH09.0101 A.2.13. 7-43c All Indicated Changes are discussed in RAI 2-1

1VIP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.2.13. 7.5 References I

1. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section Ill, Subsection NB, see Chapter A.2, Section A.2.1.2.1 for applicable editions.

2. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section III, Appendices, see Chapter A.2, Section A.2.1.2.1 for applicable editions.

3. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section II, Part D, see Chapter A.2, Section A.2.1.2.1 for applicable editions.

4. ANSYS Computer Code and User's Manual, Release lOAl(dn_di.~z~dse)fIJ

5. ISG-15, Rev. 0, "Materials Evaluation."

6. R.A. Robinson, et. al., "A survey of Strain-Rate Effects for some common Structural Materials used in Radioactive Material Packaging and Transportation Systems", Report BMl-1954, August 1976, Battelle Columbus Laboratories.

7. Tietz, T. E., "Determination of the Mechanical Properties of a High Purity Lead and a 0.058

% Copper-Lead Alloy," WADC Technical Report 57-695, ASTIA Document No. 151165, Stanford Research Institute, Menlo Park, CA, April 1958.

8. H.J. Rack, G.A. Knorovsky, "An Assessment of Stress-Strain Data Suitable for Finite-Element Elastic-Plastic Analysis of Shipping Containers," Sandia Laboratories, NUREG/CR-0481, SAND77-1872 R-7, 1978.

9

- A;,i~~i~~~ soa~~ o/id;~ii~~ic~T'iiniin~~;s:-;fsj.jjj; Baize;-;tzaP;e;;iif; ve~;~-i coJ~::s~*~tion:1

, -. -f!Il; Subsectfon N_F,~ see ChapterJ~2; Sectio~ A.2: 1.2.1 for applicable editions)'.------ -- --------- -- - --** "... ""'--*'*-* *".,-.,,'--**~r"***~l.,-~.._ ___ +* ~-

• • -~*

-~~--,- -~--- "" *'** -

• -- -****-----"'-*r*-*---*""* -

~-*--- * * --* --~*~*-****..-......,......---,. +

• " -~.---......J NUH09.0101 A.2.13.7-44 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13.7-4

---

-- --- - ------- -- --- -- --- l-\\:llowabie-stres-sValues-fot Stainless-steel and-carbon Steel at soo-°Flf.or(IJ$Cs( -- -- --- ------- -------

Loading Stress Stress Allowable Condition Cate~ory Criteria r11 Material Stress (ksi) Membrane Stress, Sm Stainless Steel 17.5 Pm A36 19.3 Normal Membrane+ Stainless Steel 26.3 Conditions, Bending Stress, 1.5 Sm A36 29.0 Elastic Pm+Pb Analysis Primary+ Stainless Steel 52.5 Secondary Stress, 3 Sm A36 57.9 Pm+Pb+O Accident Membrane Stress, min of Stainless Steel 42.0 Pm (2.4 Sm, 0.7 Su) A36 40.6 Conditions, Membrane+ Stainless Steel 63.0 Elastic-Bending Stress, min of Analysis Pm+Pb (3.6 Sm, 1.0 Su) A36 58.0 Membrane Stress, max.of Stainless Steel 44.4 Accident 0.7 Su, Conditions, Pm Sv+ (Su-Sv) /3 A36 40.6 Elastic-Plastic Membrane+ Stainless Steel 57.1 Analysis Bending Stress, 0.9Su A36 52.2 Pm+Pb NUH09.0101 A.2.13.7-48 All Indicated Changes are discussed in RAI 2-1

l'.v1P197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13. 7-4a

---

:A.-llvwable*Stress-V-aluesforStainl-ess-Steelat-J*S-o-°FforR-WCr * * * *-* ---*-*- **-* *---- - --* - ** -

Service Stress Category Stress Criteria Allowable Stress Level (a), 150 °F (ksi) Membrane Stress, fJ1<1.0S 18.35 Pm Membrane+ Level A Bending Stress, fJ1+fJ2<1.5S 27.5 Pm+Pb Primary+ Secondary Stress, Pi+Pb< min(2 Sy, S,J 53.4 P1+Pb _ Membrane Stress, Pm<min(max(1.2 Sy,1.5 S,,J,0.7 S,J 32.04 Level D Pm (Elastic Membrane+ Analysis) Bending Stress, Pm+Pb<1.5*min(max(1.2 Sy,1.5 S,,J,0.7 S,J 48.06 Pm+Pb Membrane Stress, Pm<max(Sy+(Su-Sy)/3,0. 7 S,J 51.1 Leve!D Pm (Plastic Membrane+ Analysis) Bending Stress, Pm (or PJ+Pb< 0.9 Su 65.7 Pm+Pb NUH09.0101 A.2.13. 7-48a All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13.7-5

---

-*** *- --- ** -- **-- ---- *** *------- * ----* * ---*--*---- -A:Uowable-Weld-Stresses{t¢i_::.:§§e.~+-------*- ----* ________ : __ ----* -------,-*-*--- -

Allowable Stress Value at Service Level Stress Region/Category Stress Criteria 500 °F fksil A Weld Stress away from Impact Zone 0.8 fl.5 Sm] 21.0 (Normal Condition Weld Stress in local area at Impact Zone 0.8 [3.0 SmJ 42.0 of Transport) D Primary Membrane Stress, P., 0.8 [Max( 0. 7 Su, Sv + (Su - Sv)/3)] 35.5 (Hypothetical Primary Membrane ( or Local Membrane) 0.8 [0.9 SJ 45.7 Accident Condition) + Bending Stress, P., (or P,) + Pb NUH09.0101 A.2.13.7-49 All Indicated Changes are discussed in RAI 2-1

l\\1Pl97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.2.13. 7-5a

---

------ ------- ------ --- **- *------ ----------** AJlowaoleWeldStresses at 150-°Ffor RWCs ------- - *- ----------------- --------- *- *------- -

Stress Allowable Stress Service Level Stress Criteria Value at 150 °F Region/Category [ksi] A Weld Metal 0.3 Fu 24.0 (Normal Condition of Base Metal 0.4 Sy 10.68 Transport) D Weld Metal 0.3 Fux min(2,l.167SjSv) 48.0 (Hypothetical Accident Base Metal . 0.4 Syx min(2,l.167SjSy) 21.36 Condition) NUH09.0101 A.2.13.7-49a All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I FE MODEL-TOP END FE MODEL-BOTTOM END Figure A. 2.13. 7-8a Group 5 RWCs Typical 3D Finite Element Model Mesh NUH09.0101 A.2.13. 7-57a All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report A.2.13.9.2 Natural Frequencies during End Drop Canister Shell Rev. 18D, 12/18 I The canister frequency is calculated using the following equation from Reference [2]. Natural frequency is calculated by,

where, where, g= 386.4

~=Max. Deflection ratio= (W*L)/(A *E), W = weight of canister - weight of bottom shield plug, L = length of cavity, E = Young's modulus, A = Cross sectional area The natural frequencies for the entire set of canister shells for the DSC and RWC are calculated and summarized in Table A.2.13.9-1; a sample calculation for the bounding canister shell, i.e., the 32PT canister, is presented below: The maximum temperature in the canister shell for normal transport condition is less than 500 °P (Chapter A.3). However, the canister material property is conservatively taken at 500 °F. The canister shell is constructed from SA-240 Type 304, which has a modulus of elasticity of 25.8 x 106 psi at 500 °P. The cavity length of the canister between top and bottom assemblies is 175.6 in. W = 24,393 - 5,366 = 19,027 lbs A= (n/4 )*(67.192 -66.192) = 104.77 in2 ~ = WL/AE = (19,027*175.6)/(104.77*25.8 x 106) = 0.001236 I, = _1 g = _1 386.4 = 88.9 Hz ( ) 1/2 ( )1/2 1 21f ~ 21f 0.001236 Basket The fundamental natural frequency of a simply supported basket structure under axial vibration simplifies to that of a uniform beam axially free at both ends, for all baskets except for the 24PT4 basket. The fundamental natural frequency of a uniform beam free at both ends under longitudinal load vibration is as follows [2]: NUH09.0101 A.2.13.9-2 All Indicated Changes are discussed in RAI 2-1

l\\t1P197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 are* 60.8 Hz, 229.1 Hz, and 87.3 Hz for the spacer disc, guide sleeve, and support rod assemblies, __... -*- __ ---*- _. respecth1.:elr... TheJowestmode_for.the.spacer.disds_shown.in.Eigure.A.2.13.. 9-~J... -----....... ____..... __... - ---- --- - - Fuel Since fuel end drops are evaluated using dynamic analyses, DLF for the fuel claddings is not needed and thus the natural frequency for the fuel is not calculated. A.2.13.9.3 Natural Frequencies during Side Drop Canister Shell The fundamental natural frequency of the canister is assumed to be caused by a cylindrical shell ovalling mode. The fundamental natural frequency of the canister shell ovalling (Radial-Axial) mode is determined assuming the cylindrical shell is simply supported without axial constraints. The natural frequency of the cylindrical shell ovalling mode is given by the following ([2], p. 305, Table 12-2, Frame 5).

where, E = Young's modulus R = Average shell radius v = Poisson's ratio

µ=Mass density, (lbm.in.-3). For the fundamental mode, i = 2 and j = l. Av=

where, K1-v2)(j1iRI L)4 + (h2 /l2R2)[i2+(}1iR/ L)2 ]4 r2 (jm?.I L)2 + i 2

v = Poisson's ratio R = Average shell radius L = Cavity length h = thickness The natural frequency for the entire set of canister shells/or the DSC and RWC is calculated and summarized in Table A.2.13.9-3; a sample calculation for the bounding canister shell, i.e., the 24PT4 canister, is presented below: The maximum temperature in the canister shell for normal transport condition is less than 500 °F (Chapter A.3). However, the canister material property is conservatively taken at 500 °F. The canister shell is constructed from SA-240 Type 304, which has a modulus of elasticity of25.8 x NUH09.0101 A.2.13.9-4 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I The natural :frequencies for the fuel claddings are summarized in Table A.2.13.9-5 and the lowest _ __ _ _ _ _ __ __ _ ___ frequency_for_the_ hounding_ fueLcladding_is_sho.wn__inEigure_A.2.B.9::_4.__ ___ _ _____ __________ ___ _ _ ____ __ _ _ _ __ __ -_ A.2.13.9.4 Dynamic Load Factor Calculations ANSYS transient dynamic analysis was performed using a finite element model consisting of a damped spring oscillator, COMBIN14, and structural mass, MASS21, elements to calculate DLFs for I' and 30' end and side drops. The finite element model is shown in Figure A.2.13.9-

5. The acceleration time history of the I ft and 30 ft end and side drops are applied to the mass element and DLF is calculated by:

DLF = Umax dynamic/ Umax static

where, Umax static - calculated by max g load x mass / stiffness Umax dynamic - calculated by AN SYS Unit mass is assumed for all analyses, and the stiffness and damping (7% damping is used) are calculated based on the desired :frequency of the spring using the following equations:

k= m(2ef)2

where, m-mass f-desired frequency C =2s.J(km)

where,

( - damping ratio k-stiffness m-mass DLFs are calculated for each drop condition for a frequency range from 5 to 200 Hz and are shown in Figure A.2.13.9-6. The DLFs for all baskets, DSC and RWC canisters, and fuel claddings are summarized in Table A.2.13.9-6 and Table A.2.13.9-7. The LSDYNA analyses in Appendix A.2.13.12 have been updated in Rev. 7, however since the effect on acceleration time histories is minimal, the results for the DLF are unaffected. A.2.13.9.5 References I. ANSYS Computer Code and User's Manual, Release, 8.1 and IOAI.

2. Blevins, Robert D., "Formulas for Natural Frequency and Mode Shape," Krieger Publishing Company, Florida, 1995.

NUH09.0I01 A.2.13.9-6 All Indicated Changes are discussed in RAI 2-1

MP197 Transportation Packaging Safety Analysis Report Table A.2.13.9-1 Lowest Natural Frequencies of the Canister Shell during End Drop Canisters Cavity Length of Canister (in) Weight of Entire Canister (lbs) Weight of Bottom Shield Plug (lb) Outer Diameter of Canister (in) Inner Diameter of Canister (in) Cross Sectional Area (in 2 ) Young's Modulus @500DegF Maximum Static Vertical Deflection I:!. (in) Natural Frequency of Canister (End Drop), f1 Hz Notes: 61BT/,.,. 69BTH 61BTH 179.5 178.4 22329 24594 3957 7805 67.25 69.75 66.25 68.75 104.86425 108.791 2.58E+07 2.58E+07 0.001219 0.00106 89.52 95.67 24PTH 24PT4 175.1 180.2 22013 15626 3927 4211 67.19 67.19 66.19 65.94 104.769 130.717 2.58E+07 2.58E+07 0.00117 0.000610 91.31 126.55

1. The Young's Modulus of the RWC is considered to be at 150 °F.

NUH09.0I01 32PT 175.6 24393 5366 67.19 66.19 104.769 2.58E+07 0.00123 88.90 A.2.13.9-7 32PTH/ 32PTH1 181.75 24048 7203 69.75 68.75 108.791 2.58E+07 0.00109 94.63 37PTH 171.63 24388 7758 69.75 68.75 108.79 2.58E+07 0.0010 98.01 All Indicated Changes are discussed in RAI 2-1 RWC-W 167.3 32,306 3,856 67.19 64.69 258.94 2.78E+O'f1J 0.0006613 121.62 RWC-B 167.3 37,198 3,856 67.19 63.69 359.77 2.78E+O'f1J 0.0005577 132.422 Rev. 18D, 'i/. 2/18 RWC-Dl) I i 183.251 I i 32,2701 I 3,471 i I 67.25 i \\ I 63.75; I 360.l I I 2.78E+otJ I i 0.00052'1!2 I 136.2031 I I I I I I I I I I I I I I I I I I

MP197 Transportation Packaging Safety Analysis Report i Rev. l BD, f 2/18 Table A.2.13.9-3 Lowest Natural Frequencies of the Canister Shell during Side Drop 61BT/ 32PTH/ Canister 618TH 698TH 24PTH 24PT4 32PT 32PTH1 37PTH RWC-W RWC-B RWC-DD v - Poisson's ratio 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 - i For fundamental 2 2 2 2 2 2 2 2 2 2 mode i =2 j For fundamental 1 1 1 1 1 1 1 1 1 1 modej =1 µ - Mass Density 0.0007 0.0008 0.0007 0.0008 0.0007 0.0008 0.0008 0.00075 (J.00075 0.00075 2.58E+O 2.58E+O 2.58E+O 2.58E+07 2.58E+07 2.58E+07 2.58E+07 2. 78E+071J 2. 78E+Of1J 2. 78E+071J E -500F 7 7 7 h - thickness 0.5 0.5 0.5 0.625 0.5 0.5 0.5 1.25 1.75 1.75 R - Average Shell 33.345 34.625 33.345 33.2825 33.345 34.625 34.625 32.97 32.72 32.75 Radius L - Cavity Length 179.5 178.4 175.6 180.2 175.6 181.75 171.63 167.3 167.3 183.25 0.0764 0.082461 0.07937 0.076352 0.079377 0.079732 0.088413 0.096 0.106 0.096 .,\\;; 7 Ovalling 71.29 73.78 74.35 71.07 74.35 71.34 79.11 93.68 104.44 94.44 Frequencv Note: I. The Young's Modulus of the RWC is considered to be at 150 °F. NUH09.0101 A.2.13.9-9 All Indicated Changes are discussed in RAI 2-1

l'v1P197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I

• Table A.2.13.9-6

---

- --- - - -- --- Dynarn1c*Load Factor Results Summary ______ *----- --- -------- *----------

61BT/ 32PTH/ RWC RWC 61BTH 69BTH 24PTH 24PT4C1l 32PT 32PTH1 37PTH -W -B sd-1.36 1 ft Basket gs-1.00 End Drop 1.00 1.00 1.00 sr-1.33 1.00 1.00 1.00 n/a n/a 1 ft Canister End Drop 1.32 1.31 1.32 1.18 1.33 1.31 1.30 1.20 1.15 sd-1.31 30 ft Basket gs-1.00 End Drop 1.25 1.25 1.25 sr-1.17 1.25 1.25 1.25 n/a n/a 3 0 ft Canister End Drop 1.17 1.16 1.16 1.17 1.17 1.16 1.15 1.16 1.21 sd-1.05 1 ft Basket gs-1.00 Side Drop 1.06 1.06 1.13 sr-1.04 1.27 1.17 1.25 n!a n/a 1 ft Canister Side Drop 1.27 1.29 1.30 1.26 1.30 1.26 1.31 1.27 1.18 sd-1.00 30 ft Basket gs-1.00 Side Drop 1.00 1.014 1.014 sr-1.01 1.04 1.01 1.00 n/a n/a 3 0 ft Canister Side Drop 1.01 1.00 - 1.00 1.01 1.00 1.01 1.02 1.01 1.01 Note: (General) For component frequencies >200 Hz, the frequency of200 Hz is conservatively used to calculate the DLF. (1) sd-spacer disk; gs-guide sleeve; sr-support rod NUH09.0101 A.2.13.9-12 All Indicated Changes are discussed in RAI 2-1 RWC -DD n/a 1.12 n/a 1.22 n/a 1.26 n/a 1.01

MP197 Transportation Packaging Safety Analysis Report NUH09.0101 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 A.2.13.14-7 All Indicated Changes are discussed in Enclosure 3, Item 1 Rev. 18D, 12/18 I

MP197 Transportation Packaging Safety Analysis Report Chapter A.5 Shielding Evaluation TABLE OF CONTENTS Rev. 18D, 12/18 I A.5.1 Description of the Shielding Design............................................................................... A.5-1 b A.5.1.1 Package Design Features........................................................................................... A.5-lb A.5.1.2 Codes and Standards.................................................................................................... A.5-2 A.5.1.3 Summary Table of Maximum Radiation Levels.......................................................... A.5-2 A.5.2 Source Specification......................................................................................................... A.5-3 A.5.2.1 Gamma Source............................................................................................................. A.5-4 A.5.2.2 Neutron Source............................................................................................................ A.5-7 A.5.2.3 Axial Source Distribution............................................................................................ A.5-8 A.5.2.4 Axial Blankets............................................................................................................ A.5-10 A.5.2.5 HAC Sources..........,.................................................................................................. A.5-11 A.5.3 Model Specification....................... *................................................................................. A.5-12 A.5.3.1 Configuration of Source and Shielding...................................................................... A.5-12 A.5.3.2 Material Properties..................................................................................................... A.5-14 A.5.4 Evaluation....................................................................................................................... A.5-15 A.5.4.1 Methods..................................................................................................................... A.5-15 A.5.4.2 Flux-to-Dose-Rate Conversion................................................................................ A.5-30a A.5.4.3 Radiation Levels........................................................................................................ A.5-31 A.5.5 Appendix......................................................................................................................... A.5-33 A.5.5.1 VYAL-B Mixing and Installation.............................................................................. A.5-33 A.5.5.2 Fuel Qualification Results.......................................................................................... A.5-33 A.5.5.3 Decay Heat Restrictions................................,............................................................ A.5-35 A.5.5.4 Fission gas for BWR fuel at 70 GWd/MTU............................................................ A.5-38a A.5.5.5 Reduced Lead Thickness Assessment....................................................................... A.5-38a A.5.6 References....................................................................................................................... A.5-39 NUH09.0101 A.5-i All Indicated Changes are discussed in Enclosure 3, Item 2

I RAI 5-1 MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 Table A.5-49 MCNP Benchmark Results for HSM Model 80........................................................... A.5-SOj Table A.5-50 ORI GEN-ARP Input Parameters for the Design Basis Sources................................. A.5-80k Table A.5-51 DNcI Neutron Dose Rate Values for Various Active Fuel Compressions.................... A.5-801 Table A.5-52 "B" Parameters to Determine Additional Cooling Time for Reconfigured BWR Fuel (years)................................................................................................................ A.5-80m Table A.5-53 "B" Parameters to Determine Additional Cooling Time for Reconfigured PWR Fuel (years)................................................................................................................. A.5-80n Table A.5-54 Key Parameters of Compressed Fuel for 69BTH DSC Fuel Reconfiguration............ A.5-800 Table A.5-55 NCT Reconfigured Fuel Dose Rates for MP197HB Containing 69BTH DSC with Compressed Fuel Configuration, 41 % Void Fraction by Volume (3 0% Fuel Compaction)........................................................................................................ A.5-80p Table A.5-56 NCT Reconfigured Fuel Dose Rates for MP197HB Containing 69BTH DSC with Compressed Fuel Configuration, 17% Void Fraction by Volume (50% Fuel Compaction)........................................................................................................ A.5-80q Table A.5-57 HAC 1 Meter Dose Rates of MP197HB Containing 69BTH DSC with Compressed Fuel Configuration, 41 % Void Fraction by Volume (30% Fuel Compaction)................................................................................................................ A.5-80r Table A.5-58 HAC 1 Meter Dose Rates ofMP197HB Containing 69BTH DSC with Compressed Fuel Configuration, 17% Void Fraction by Volume (50% Fuel Compaction)................................................................................................................ A.5-80s Table A.5-59 69BTH NCT 17% Compressed Fuel - Neutron and Secondary Gamma - Burned Fuel Composition and MCNP Fission Neutron Multiplication Evaluation-Design Basis Source 2.6 wt.% U-235 I 62 GWd/MTU......................... A.5-80t Ta e A.5-60 69BTH NCT 17% Compressed Fuel - Neutron and Secondary Gamma - Burned Fuel Composition and MCNP Fission Neutron Multiplication Evaluation - High Enriched and High Bumup Fuel (5.0 wt. % U-235 I 70 GWd/MTU / 12 yrs CT)............................................................................................... A.5-80t I Ii Table A.5-61 Eauivalent Activitv Limits as a Function of Enerf!V...................................................., ~ nn Table A.5-62 Equivalent Activity Limits as a Function of Energy-MP 197HB Unit 01 - Reduced Lead Thickness............................................................................................. A.5-80v Table A.5-63 Reduced Lead Evaluation - 69BTH DSC................................................................... A.5-80w See Enclosure 3, Item 2 NUH09.0101 A.5-iv All Indicated Changes are discussed as indicated above

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I [ MP J 97HB Unit OJ is the same as the MP J 97HB with the exception of localized reduced lead thickness on the side of the cask body. The shielding performance of MP J97HB Unit OJ is not expected to be affected for the authorized spent fuel content as shown in Section A.5.5.5.J. 70,000 Ci of Co-60 or equivalent is demonstrated to be acceptable for a uniform lead thickness of 2. 77 inches instead of 3. 00 inches nominal thickness. A.5.1 Description of the Shielding Design The MP197HB cask is designed to transport one of several NUHOMS DSCs loaded with spent fuel assemblies or dry irradiated and/or contaminated non-fuel bearing solid materials in a radioactive waste canister (RWC) in accordance with the requirements of the 10 CPR 71. The authorized contents acceptable for transport are described in Chapter A.1, Section A.1.2.3, including appendices A.1.4.1 through A.1.4.9A. A complete list of the NUHOMS DSCs authorized for transport is provided in Chapter A.1, Section A.1.2.3.1. Chapter A.1, Section A.1.2.3.2 (also in Appendix A.l.4.9A) provides a description of the irradiated and/or contaminated non-fuel bearing solid materials authorized for transport in the RWC as well as its respective physical dimensions. Radiological sources used for the calculation of the dose rates presented in this chapter are determined through ranking using the response function methodology to develop the fuel qualification tables (FQT). Response function results are compared with direct MCNP analysis using a discrete MP197HB transportation package model as described in Section A.5.4.1.2.3. By definition of the FQTs, the minimum cooling times are determined so that the maximum NCT dose rates for intact fuel at 2 m from the side of the vehicle are :S 8.2 mrem/hr. For fuel in the peripheral basket locations, additional cooling time is needed for some bumup, enrichment, and cooling time (BECT) combinations due to fuel reconfiguration, as defined using the methodology in Section A.5.4.1.3.3. Further discussion of the fuel qualification methodology is contained in Section A.5.4.1.3 and FQT results are discussed in Section A.5.5.2. NUH09.0101 A.5-lb All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 A.5.1.1 Package Design Features Shielding for the MP 197HB transportation package is provided mainly by the cask body. Shielding/or gamma radiation is provided by the lead and stainless steel shells that comprise the cask wall. Nominal lead thickness is 3. 00 inches for the MP l 97HB transportation package. For the neutron shielding, a borated VYAL-B resin compound surrounds the cask body radially. Gamma shielding in the cask ends is provided by the steel top and bottom assemblies of the transportation cask and axial ends of the DSCs. Additional shielding is provided by the steel outer shell surrounding the resin layer, the steel and aluminum structure of the fuel basket and optional heat dissipation fins surrounding the cask side between impact limiters. For transport, wood filled impact limiters are installed on either end of the cask and provide additional shielding for the ends and some radial shielding for the areas at either end of the radial neutron shield. Important-to-shielding dimensions are shown in Table A.5-4. NUH09.0101 A.5-lc All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 Hypothetical accident condition (HAC) dose rates are calculated 1 m from the surface of the cask body. No credit is taken for the neutron shield or impact limiters. Both intact and reconfigured fuel are considered. The maximum radiation dose rates for HAC are shown in Table A.5-2 [ ] and Table A.5-2a [ ] Dose rates for RWC are provided in Table A.5-34. This table contains both NCT and HAC dose rate results. Compared to spent fuel, RWC dose rates are low. A.5.2 Source Specification There are five principal sources of radiation associated with transport of spent nuclear fuel that are of concern for radiation protection.

1. Primary gamma radiation from spent fuel.

2. Primary neutron radiation from spent fuel (both alpha-n reactions and spont3:neous fission).

3. Gamma radiation from activated fuel structural materials and fuel inserts.

4. Capture gamma radiation produced by attenuation of neutrons by shielding material of the cask.

5. Neutrons produced by sub-critical multiplication in the fuel.

There are three source configurations used in the evaluation of the shielding performance of the MP197HB transportation package. These configurations are selected because of their respective bounding parameters on all authorized contents. The bounding configurations are as follows: 90,000 Ci of Co-60 or equivalent in the RWC, (equivalent activities for sources not entirely Co-60 are discussed in Section A.5.2.1.5 and shown in Table A.5-61), except for MP197HB Unit OJ where the limit is 70,000 Ci ofCo-60 or equivalent due to a reduction on lead thickness for the as-built condition (equivalent activities for sources not entirely Co-60 are discussed in Section A.5.2.1.5 and shown in Table A.5-62), 69 GE-2,3 7x7 Type G2A BWR spent fuel assemblies in the 69BTH DSC, and 37 B&W 15x15 Mark B-10 PWR spent fuel assemblies in the 37PTH DSC. For the spent fuel assemblies listed, design basis sources which encompass the allowable burnup and enrichment combinations for the authorized contents are developed in the following subsections. The spent fuel assembly types are selected as bounding mainly because their respective initial uranium loading bounds all others. NUH09.0101 A.5-3 All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 For the 37PTH and 69BTH DSCs in the MP197HB transportation package, design basis BWR and PWR spent fuel sources are developed based on a bounding assembly average bumup, initial enrichment, and cooling time. These parameters are selected based on the fuel qualification method discussed in Section A.5.4.1.3. The B&W 15x15 Mark BIO and the GE-2, 3 7x7 Type G2A fuel assemblies contain the maximum heavy metal weight for their type, nearly 490 and 198 kgU, respectively. They result in bounding neutron and gamma source terms for PWR and BWR type of assemblies, respectively. Therefore, B& W l 5xl 5 Mark B 10 and the GE 2, 3 7x7 are evaluated as the design basis (DB) PWR and BWR fuel assembly (FA) in the shielding evaluation of MP197HB transportation package, respectively. The fuel assembly hardware for the GE 7x7 fuel assembly is bounding as it contains the maximum amount of steel and inconel than any other BWR fuel design. For the PWR fuel assembly designs with cooling times greater than 15 years, the contribution from the fuel assembly hardware becomes less important due to substantial decay of the Co-60 source from hardware irradiation. NUH09.0101 A.5-3a

MPI97 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.5.2.1.4 Control Components The spent fuel payload consists of various DSCs with PWR fuel assemblies and associated control components (CCs) and is specified in Chapter A.I, Appendix A.1.4.1 through A.1.4.6. For the PWR fuel assemblies, the various authorized CCs are listed in the above appendices. These include PWR burnable poison rod assemblies (BPRAs), thimble plug assemblies, control rod assemblies, control rod cluster assemblies, axial power shaping rods, orifice rod assemblies, vibration suppression inserts, neutron source assemblies, and neutron sources. The CCs are typically solid or hollow rods of stainless steel or Zircaloy containing neutron absorbing or neutron source materials. Typically, the source term from these CCs is dominated by the Co-60 spectrum. Therefore, a separate material composition and irradiation history is not necessary for characterizing all of these CCs. Radiological source in Table A.5-18 bounds any CC authorized for loading. The source in this table is referred to as design basis CC source. Guidelines for adjustment ofFQT cooling times due to presence of DB CC sources are provided in Section A.5.5.2.1. DB PWRFA BPRAs with burnup between 36,000 MWd/MTU and 45,000 MWd/MTU are bounded by the design basis CC source after 8 years decay. All other BPRAs irradiated between 36,000 MWd/MTU and 45,000 MWd/MTU would require 13 years of decay to be bounded by the design basis CC source. All other CCs would need to be examined on a case-by-case basis. Combinations ofradiological sources due DB PWR assembly and the DB CC source result in bounding dose rates when evaluating shielding performance of MP197HB transportation package loaded with DSCs containing PWR F As with DB CC sources. A.5.2.1.5 RWC The NUHOMS-MP197HB is designed to transport a payload of 56.0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The safety analysis of the cask takes no credit for the containment provided by the RWC. The quantity of radioactive material is limited to a maximum of 90,000 Ci of Co-60 or equivalent with the exception of MP 197HB Unit 01 where the content is limited to a maximum o/70,000 Ci ofCo-60 or equivalent. A list of typical components and their associated activities is shown in Section A.1.4.9A.3. Co-60 emits two photons per disintegration, one at 1.17 MeV and one at 1.33 MeV. The decay heat load of the radioactive material is expected to be less than 5 kW, which is well below the 26 kW limit for the cask. The limiting configuration from Table A. 5-34, 2 meters from the package side in NCT, is employed to determine the gamma activity limits at each energy group. NUH09.010I A.5-6 All Indicated Changes are discussed in RAI 5-1 and Enclosure 3, Item 2

RAI 5-1 MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18

• 1 Energy-dependent activities are developed using the 18-energy-group, 0. 6-10 Me V, response functions at 2 meters from the package side in NCT. Assumptions associated with the MCNP model employed/or developing the response functions are identical to those employed/or calculating the dose rates shown in Table A.5-34, in particular the source is uniformly distributed throughout an assumed waste volume with a 66. 0-inch diameter and a 168-inch height, and the density of the source material is 1.0 glee. The results shown in Table A.5-61 are the limits on payload gamma activity as function of gamma energy (90,000 Ci Co-60 or equivalent). The response functions are developed with one angular bin, circumferential average tally, while dose rate reported in Table A.5-34 are computed with 71 angular bins.

. Activity limits for gamma energy emissions other than Co-60 are determined as the equivalent activities per gamma energy to an activity of 90,000 Ci of Co-60 resulting in a dose rate limit of

7. 71 mrem/hr. Note that a significant reduction in activity is observed for emitters with higher gamma energy than Co-60 gammas. This is consistent with the attenuation ability of materials for gamma rays. Higher source energies require more material for equivalent shielding.

For sources that are n_ot entirely Co-60, the following equation is used: L S;(E) ~1

Activity Limit;(E) where S;(E) is the source strength in i and ActivityLimif; (E) is the corresponding maximum emission for the each energy of the radionuclide gamma emission. The energy should be rounded up to the next higher energy found in Table A.5-61.

For example,for a content ofCo-60 and Cs-137 mixture, the allowed activities ofCo-60 and Cs-137 are defined as: I 016

• I 15 1s Yco-60 1.33xl

+ Yco-60 4.44x10 + Ycs-137/ I.OlxlO :SI Where Yco-60 is the activity of Co-60 and Ycs-137 is the activity of Cs-13 7. 1 ne eJJecr OJ me reaucea teaa rmc,mess on me energy-aepenaenr acnviry umirs zs evatuarea Jor MP 197HB Unit OJ. The results shown in Table A.5-62 are the limits on payload gamma activity for MP 197HB Unit 01 as function of gamma energy for 70,000 Ci Co-60 or equivalent. The under-thickness lead poured for MP 197HB Unit OJ is modeled uniformly at 2. 77 inches lead poured thickness throughout side of the cask body. The limiting dose rate, 2 m side (Table A.5-34),for the maximum RWC content o/90,000 Ci is evaluated for MP 197HB Unit OJ with uniform 2. 77 inches lead thickness throughout side of the cask body. The 2 m side dose rate resulting from the evaluation, 12.11 mrem/hr, exceeds the NCT regulatory limit, IO mrem/hr. The maximum allowable content is reduced to 70,000 Ci of Co-60/or MP197HB Unit OJ in order to achieve 9.41 mrem/hr (12.11 x 70000/90000). See Enclosure 3, Item 1 NUH09.0101 A.5-6a All Indicated Changes are discussed as indicated above

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 The limits on payload gamma activity as function of gamma energy are developed using the methodology described above including 18-energy group response function at 2 m from the package side in NCT for MP 19 7 HB Unit O 1 with uniform 2. 77 inches lead thickness throughout side of the cask body. Activity limits for gamma energy emissions other than Co-60 are determined as the equivalent activities per gamma energy to an activity of 70,000 Ci of Co-60 resulting in a dose rate limit of 8. 71 mrem/hr, Table A.5-62. The difference in dose rates estimated at 2m using the response function and explicitly evaluated by scaling the results for 90,000 Ci are due to difference in the tally option used in the MCNP calculations. The response function are developed with one angular bin, circumferential average tally, while dose rate are computed with 71 angular bins. NUH09.0101 A.5-6b All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.5.5.4 Fission gas for BWR fuel at 70 GWd/MTU The bounding quantity of gases released as a result of irradiation for the "generic" GE 7x7 BWR fuel assembly is evaluated. Since the amount of released fission gases is conservative on the basis of lowest enrichment for a given bumup, the quantity of gases released is evaluated for a bumup and enrichment combination of 70 GWd/MTU and 3.70 wt% initial U-235, per BWR FQT in Tables A.1.4.9-4 to A.1.4.9-Sa. The total bounding amount of moles of gases released as a result of irradiation for one fuel assembly is presented in Table A.5-35; data presented in the table corresponds to 3.0 and 5.0 years of cooling time and 0.198 MTU. The amount of gas produced for a "generic" BWR fuel assembly, 0.198 MTU, at 70 GWd/MTU, 3.70 wt% initial U-235 is 23.0 g-moles. A.5.5.5 Reduced Lead Thickness Assessment The gamma shield nominal lead thickness is 3.00 inches; duringfabrication, and prior to the installation of the neutron shield, gamma scanning is used to verify the integrity of the poured lead shielding and a minimum thickness is confirmed by comparison of gamma scan results to a calibration block consisting of a known thickness of lead between steel plates of nominal thickness the same as used in the cask fabrication. This section provides an assessment of localized reduced lead thickness for allowable spent fuel content and maximum allowable RWC content. Reduced lead thickness occurs at different areas of the cask. Reduced lead thickness in the vicinity of the shear key is on portions of the cask side that are facing the ground during transportation. Those portions would be either partly or entirely shielded by parts of a transportation trailer such as frame bars, wheels, gear boxes, etc. A.5.5.5.1 Spent Fuel Content Assessment The impact of localized under-thickness lead is illustrated with the following evaluation considering three grooves on side of the cask representing segments of the lead shielding with 3 mm thickness below the nominal value. The segments extend axially throughout an entire side of the cask body with 30-inch width. Two of the segments are located at angular coordinates above the trunnions and one segment is at the shear key position. The evaluation is performed for the 69BTH DSC for the bounding NCT and HAC configurations shown respectively in Table A.5-la and Table A.5-2. The dose rates, at 2 min NCT and 1 min HAC, are summarized in Table A.5-63. The impact on dose rates near the cask with reduced lead shielding thickness for gamma radiation is the most pronounced when the contribution to the total NCT dose rate due to gamma radiation is the largest. NUH09.0101 A.5-38a All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 As observed from Table A.5-63, the effect of the reduced lead thickness on the limiting dose rate for the spent fuel contents is not significant. This is an expected result as the neutron source is the predominate contributor to the limiting dose rate. Figure A.5-21 and Figure A.5-22 illustrate the fraction of total dose rate due to neutron radiation source for respectively 37PTH DSC and 69BTH DSC; one can observe that fractions of the total NCT dose rates due to gamma radiation are greater than 50% only for burn-ups that are about 45 GWD/MTU and below. The transportation FQT dose rates for all bounding authorized contents in 37PTH DSC and 69BTH DSC are shown in Table A.5-26 and Table A.5-

30. NCT dose rates, shown in Table A.5-26 and Table A.5-30, at 2 mfor burnup about 45 GWDIMTU and below are well below 7.8 mremlhr. This is due to the fact that cooling times in FQTs AppendixA.1.4.6/or 37PTH DSC andAppendixA.1.4.9 for 69BTH DSC are determined in order not to exceed regulatory limits on dose rates and satisfy criticality safety requirements.

From the evaluation and observations above, localized reduced lead thickness condition does not make the bounding dose rates exceed the regulatory limits in NCT and HAC conditions since the margins to the regulatory limits are significant for spent fuels with burnups about 45 GWDIMTU and below (dominant gamma radiation burnup range) and neutron radiation is significantly dominant for high burnup fuels. A.5.5.5.2 RWC Content Assessment Authorized RWC content is entirely due to gamma radiation, and localized reduced lead thickness condition increases the bounding NCT dose rate when the maximum allowable content is applied. The maximum allowable content is reduced in order to not exceed the NCT regulatory limits. NUH09.0101 A.5-38b All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.5-61 Equivalent Activity Limits as a Function of Energy Energy

Response

Relative Activity Dose Rate Equivalent (MeV) (mremlhlrls) Uncertainty (rls) (mremlh) Activity (rls/2J 0.6 7.98E-20 0.008 9.66E+19 0.8 7.61E-18 0.070 J.OJE+l8 1 l.17E-16 0.027 6.58E+J6 1.1732 5.78E-16 0.014 3.33E+ J 5(1J J.93E+OO 1.33E+16 1.3325 J.74E-15 0.009 3.33E+ 15(1) 5.78E+OO 4.44E+l5 1.5 4.19E-15 0.007 J.84E+l5 1.75 J.JJE-14 0.005 6.92E+14 2 2.27E-14 0.004 3.39E+l4 2.5 6.05E-14 0.003 1.27E+14 3 1.JJE-13 0.003 6.83E+J3 3.5 l.72E-13 0.003 4.47E+l3 4 2.29E-13 0.003 3.36E+13 4.5 2.85E-13 0.003 2.71E+J3 5 3.27E-13 0.003 2.36E+J3 6 3.96E-13 0.003 1.95E+l3 8 4.65E-13 0.003 J.66E+l3 JO 4.96E-13 0.003 1.55E+J3 Total Dose Rate: 7.71 (J) Activity limit corresponding to 90,000 Ci ofCo-60 resulting in 7. 71 mrem/hr (using response functions at 1.1732 MeVand 1.3325 MeV). (2) Equivalent activity limits per energy for contents other than Co-60 determined using 7:71 mrem/hr and response functions at energy groups shown in the first column. Equivalent Activity (i) = 7. 71 I Response function at Energy (i). NUH09.0101 A.5-80u All Indicated Changes are discussed in RAI 5-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 Table A.5-62 Equivalent Activity Limits as a Function of Energy-MP 197HB Unit OJ - Reduced Lead Thickness Energy

Response

Relative Activity (yls) Dose Rate Equivalent (MeV) (mremlh/yls) Uncertainty (mremlh) Activity (y/s) (ZJ 0.60 1.63E-19 0.008 5.36E+l9 0.80 1.44E-17 0.070 6.07E+17 1.00 l.79E-16 0.027 4.86E+16 1.1732 8.59E-16 0.0.14 2.59E+ 15{l) 2.23E+OO l.OJE+l6 1.3325 2.51E-15 0.009 2.59E+15(IJ 6.49E+OO 3.48E+l5 1.50 5.87E-15 0.007 1.48E+15 1.75 1.54E-14 0.005 5.67E+l4 2.00 3.04E-14 0.004 2.86E+14 2.50 8.00E-14 0.003 1.09E+14 3.00 1.49E-13 0.003 5.85E+J3 3.50 2.26E-13 0.003 3.86E+J3 4.00 3.00E-13 0.003 2.90E+J3 4.50 3.73E-13 0.003 2.34E+l3 5.00 4.29E-13 0.003 2.03E+J3 6.00 5.19E-13 0.003 1.68E+J3 8.00 6.19E-13 0.003 1.41E+J3 10.00 6.66E-13 0.003 1.31E+J3 Total Dose Rate: 8.71 (I) Activity limit corresponding to 70,000 Ci ofCo-60 resulting in 8. 71 mrem/hr (using response functions at 1.1732 MeVand 1.3325 MeV) (2) Equivalent activity limits per energy for contents other than Co-60 determined using 8. 71 mrem/hr and response functions at energy groups shown in the first column. Equivalent Activity (i) = 8. 71 I Response function at Energy (i) NUH09.0JOJ A.5-80v All Indicated Changes are discussed in RAI 5-1 and Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I NUH09.0101 Table A.5-63 Reduced Lead Evaluation - 69BTH DSC Reduced Lead Thickness - 3 mm deep, 30" wide, full length grooves in lead Component of Dose NCT Dose rate at 2m 1 m from package Rate radial distance from Side surface of Impact Limiters, mrem/hr Gamma 0.228 +/- 0. 001 3.790 +/- 0.018 Neutron 6.588 +/- O.ll4 848.717 +/-3.870 (n,g) 1.809 +/- 0.012 2.430 +/- 0. 037 Total (I) 8.432 +/- 0.101 854.866 +/- 3.870 (1) Spatial locations of maximums of components of the total dose rate are generally different. Because of this, the maximum of the total dose rate is generally not equal to the sum of maximums of the components A.5-80w All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I . - ---Chapter. A.7-Package Operations TABLE OF CONTENTS A. 7.1 NUHOMS'-MP 197HB Package Loading................................................................... A. 7-1 A.7.1.1 NUHOMS'-MP197HB Cask Preparation/or Loading..................................... A.7-la A. 7.1.2 NUHOMS'-MP 197HB Cask Wet Loading.......................................................... A. 7-2 A. 7.1.3 NUHOMS'-MP 197HB Cask Dry Loading (Transferring a Loaded DSC or RWC from an Overpack into an MP 197HB Cask)........................................ A. 7-4a A.7.1.4 NUHOMS'-MP197HB Cask Preparation/or Transport.................................... A.7-7 A.7.2 NUHOMS'-MP197HB Package Unloading............................................................... A.7-8 A. 7.2.1 Receipt of Loaded NUHOMS'-MP 197HB Package from Carrier...................... A. 7-8 A. 7.2.2 Removal of Contents from NUHOMS'-MP 197HB Cask................................... A. 7-8a A. 7.3. Preparation of Empty Package for Transport........................................................... A. 7-11 A. 7. 4 Other Operations....................................................................................................... A. 7-11 A. 7.4.1 Cask Cavity Vacuum Drying and Dryness Verification Test............................. A. 7-11 A. 7. 4. 2 Pre-shipment Verification Leakage Testing of the NUHOMS'- MP 197HB Cask Containment Boundary........................................................... A. 7-12 A. 7. 5 References.................................................................................................................. A. 7-14 A.7.6 Glossary..................................................................................................................... A.7-15 A.7.7 Appendices................................................................................................................. A.7-16 TableA.7-1 Table A. 7-2a Table A. 7-2b f;.. c,, -. '\\ i Table A: 7'-2c;: \\~---*._," ~.... 2 *-*-~.1,.*.-* (Tab!~ A. 7-2rI I:.-*~ *-**--* --* __.** ' TableA.7-3 TableA.7-4 TableA.7-5 NUH09.0101 LIST OF TABLES RAI 5-1 DSC/RWC, Fuel, and Basket Spacer No nal Heights for Each Type of DSCIRWC.................................................................................................... A. 7-17 Applicable Fuel Specificationfm; arious DSCs.......................................... A. 7-18 Applicable Content Specific ion for RWC................................................... A. 7-18 {EqJJrval¢rif AcfN[ty LirflitsfqrRW<;... ~.... '...* _,,.;:.::..,,.. _.,:-;,,;.....,-,~.:-.* ~.::\\,".~;.~-.. ;A, 'j:.J&t) ftJ.*qtjvJtytfm(tifo!* -RWC:: ~ }.fP19°7f:IJJ Unit.Of.;;.:,.;,.).,.:*.~***_.';,.. :.. ;-.:,,..,.,.:;~.A. 7; J/Jb) Ap dices Containing Loading Procedu;esfor Various DSCs..............1p~-18c I Appen ices Containing Unloading Procedures for Various DSCs........... 1j.A. 7-18c I ___ d] The Unl ding Procedure which Shall Be Part of the User's Operating Procedure..................................................................................................... 7-19 RAI 5-1 and Enclosure 3, Item 2 Editorial A.7-i

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I form of the content and their maximum quantity to be loaded in any of the nine DSCs are specifiedin_TableA.7-,2a._T_ype_and_form_of_the_content-and their.maximum-quanticy to_be ______________ _ loaded in an RWC are specified in Table A.7-2b, Table A. 7-2c, or Table A. 7-2d. Equivalent Activity limits by gamma energy shown in Table A. 7-2c or Table A. 7-2d may not be interpolated in energy. The proper procedure for gamma emitter is to round source energies up to the next higher energy level in Table A. 7-2c or Table A. 7-2d. Procedures are provided in this section for (1) transport of the cask/DSC/RWC directly from the plant spent fuel pool and (2) transport of a DSC/RWC which was previously stored in a NUHOMS horizontal storage module (RSM). Section A.7.7 contains an appendix for each DSC model detailing its loading procedures. Table A.7-3 lists these appendices. A.7.1.1 NUHOMS-MP197HB Cask Preparation for Loading Procedures for preparing the cask for use after receipt at the loading site are provided in this section and are applicable for shipment ofDSCs loaded with fuel or ofRWCs loaded with dry irradiated ~.ncl/or contaminated non-fuel bearing solid materials. NUH09.0101 A.7-la All Indicated Changes are discussed in RAI 5-1

MP197 Transportation Packaging Safety Analysis Report Rev.181), J]/18 Table A.7-1 DSC/RWC, Fuel, and Basket Spacer Nominal Heights for Each Type of DSCIRWC (All dimensions are in inches) 61BTH 69BTH 24PTH 24PT4 32PT 32PTH 32PTH 32PTH1 37PTH I RWC Canister Type 61BT Tyoel Tvue2 s,, L S-LC S-100 S-125 L-100 L-125 Type! s M L s M DSC bottom 2.20 2.20 2.20 1.24 11.7 5.7 11.7 2.2 11.7 11.7 5.7 5,7 12.5 5.25 12.5 5.25 NIA 16.25 ~.o (5) snacer heil:!ht<1> I DSC top spacer (1) (1) (1) (1) (1) (1) (1) (1) (I) (1) (1) (1) (1) (1) (1) (1) (!) (!) \\1) (1) heieht Fuel spacer (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) heieht (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2)(4) (2/(4) NIA Basket spacer (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) heieht (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3)(4) (3/(4) NIA (1) I DSCIRWC top and bottom spacers can be combined to one spacer. If one spacer is used, it can be installed either on top or bottom of the DSCIRWC. Th,_e__hf!i~ht9,.(the spacer is to be determined such that the gap between the cask and DSC/RWC is below 0.5 for normal transport conditions. The specified spacer height(m_ajl_:jn~~u_i:!_~lany (2) (3) (4) (5) axial spacing provided by the internal canister sleeve components. 1 Fuel spacer can be installed either on top or bottom of the fuel assembly. The height of the fuel spacer to be determined using the formula specified in Appendix A.2.13.14, Table A.2.13.14-2 such that the gap between the fuel assemblies and the DSC is below 1.5" for normal transport conditions. Basket spacer can be installed either on top or bottom of the basket. The height of the basket spacer is to be determined such that the gap between the basket an'd the DSC is below 0.815" for normal transport conditions. Fuel and basket spacers can be combined in one spacer. Height of spacer for RWC-W and RWC-B is 11. 7 5" and height of spacer for RWC-DD is)'.~$,".! NUH09.0101 A.7-17 All Indicated Changes are discussed in Enclosure 3, Item 1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A.7-2a - - -- ---:AppHcaoleFuei-specification-for\\TariousDSCs-- Applicable Fuel Specification DSC MODEL from Chapter A.1 NUHOMS-24PT4 Tables A.1.4.1-1 and A.1.4.1-2 NUHOMS -32PT Table A.1.4.2-2 NUHOMS -24PTH Table A.1.4.3-2 NUHOMS -32PTH Table A.1.4.4-2 NUHOMS -32PTH1 Table A.1.4.5-2 NUHOMS-37PTH Table A.1.4.6-2 NUHOMS-6IBT Table A.1.4.7-2 NUHOMS-6IBTH Table A.1.4.8-2 NUHOMS-69BTH Table A.1.4.9-1 Table A.7-2b Applicable Content Specification for RWC Type and Form of Material The NUHOMS-MP197HB packaging is designed for shipment of various types of irradiated and contaminated reactor hardware. The payload will vary from shipment to shipment. Typical composition of the payload consists of the following components either individually or in combinations:

1.

BWR Control Rod Blades

2.

BWR Local Power Range Monitors (LPRMs)

3.

BWR Fuel Channels

4.

BWR Poison Curtains

5.

PWR Burnable Poison Rod Assemblies (BPRAs)

6.

PWR and BWR Reactor Vessel and Internals Decay Heat load

55kW Loading Components with high specific activity are generally placed near the center of the RWC. For each shipment, the RWC is normally filled to capacity, which prevents shifting of the contents during transport. If the RWC is not full, appropriate component spacers or shoring is used to prevent significant movement of the contents.

Maximum Quantity ofMaterial (a) For containment, the quantity ofradioactive material is limited to a maximum of per Package 8, 182 A2. The radioactive material is primarily in the form of neutron activated metals, or metal oxides in solid form. Surface contamination may also be present on the irradiated components. When a wet load procedure (i.e., in-pool) is followed for cask loading, the cask cavity and RWC are drained and dried to ensure that there are no free liquids in the package during transport. (b) The NUHOMS-MP197HB packaging is designed to transport a payload ofup to 56.0 tons of dry irradiated and/or contaminated non-fuel bearing solid materials in the RWC. The center of gravity (CG) of the loaded NUHOM:19-MP 197HB package is to be 102 +/- 4 inches from the bottom of the cask (c) The quantity of radioactive material is limited to a maximum of90,000 Ci of cobalt-60 or equivalent, except for MP 197HB Unit OJ where the limit is reduce to 70,000 Ci of cobalt-60 or equivalent. Equivalent activity limits as a.function of gamma energy for isotopes other than Co-60 are shown in Table A. 7-2c for the 90,000 Ci limit and Table A.7-2dfor the 70,000 Ci limit. The quantity of radioactive material is limited to a maximum of 90,000 Ci of cobalt-60 or equivalent, except for MP197HB Unit OJ where the limit is reduce to 70,000 Ci of cobalt-60 or equivalent. Equivalent activity limits as a.function of gamma energy for isotopes other than Co-60 are shown in Table A. 7-2c for the 90,000 Ci limit and Table A. 7-2dfor the 70,000 Ci limit. NUH09.0101 A.7-18 All Indicated Changes are discussed in RAI 5-1 and Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A. 7-2c .. ----- - --... -.. EquivalentActivity-Limits.for RWC - -- Energy Equivalent (Mer? Activity (yls) 0.6 9.66E+19 0.8 1.01E+18 1 6.58E+16 1.1732 l.33E+16 1.3325 4.44E+15 1.5 l.84E+l5 1.75 6.92E+l4 2 3.39E+l4 2.5 1.27E+14 3 6.83E+l3 3.5 4.47E+l3 4 3.36E+l3 4.5 2.71E+l3 5 2.36E+l3 6-l.95E+l3 8 l.66E+l3 10 l.55E+l3 For sources not entirely Co-60, the following equation is used: L S;(E) ~l

Activity Limit; ( E) where S; (E) is the source strength in %

and ActivityLimit; ( E) is the corresponding maximum emission for the each energy of the radionuclide gamma emission. The energy should be rounded up to the next higher energy. For example:

• For a content ofCs-137, the allowed activity ofCs-137 (0.662 MeV gamma emitter) is:

'Ycs-137/ 1.0lx1018 -::.1

Where,

'Ycs-137 is the source strength ofCs-137

• For a content ofCo-60 and Cs137 mixture, the allowed activities ofCo-60 and Cs-137 is:

'Yco-60/ l.33xl016 + Yco-60/ 4.44xl015 + Ycs-137/ l.Olxl018 -::.1 Where. 'Yco-6ois the source strength ofCo-60 and Ycs-JJ1is the source strength ofCs-137 NUH09.0101 A.7-18a All Indicated Changes are discussed in RAI 5-1

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I Table A. 7-2d

---

- ----. - _ --- - -------Activity-Limitsfor.RJif!C.-=--ME19.ZHB-Unit-OL---

Energy Equivalent (MeV) Activity (y/s) 0.6 5.36E+19 0.8 6.07E+17 1.1732 l.OJE+l6 1.3325 3.48E+l5 1.5 J.48E+l5 1.75 5.67E+14 2 2.86E+l4 2.5 l.09E+l4 3 5.85E+J3 3.5 3.86E+J3 4 2.90E+13 4.5 2.34E+l3 5 2.03E+l3 6 1.68E+l3 8 1.41E+l3 JO J.3JE+l3 For sources not entirely Co-60, the following equation is used: L S;(E) ~1

Activity Limit; (E) where S; (E) is the source strength in %

and Activity Limit; ( E). is the corresponding maximum emission for the each energy of the radionuclide gamma emission. The energy should be rounded up to the next high energy. For example: For a content of Cs-13 7, the allowed activity of Cs-13 7 (0. 662 Me V gamma emitter) is: 'Ycs-13d 6. Ox] 017 ::S 1 Where 'Ycs-137 is the source strength of Cs-137 For a content ofCo-60 and Csl37 mixture, the allowed activities ofCo-60 and Cs-137 is: 'Yco-60/ l.Olxl016 + Yco-60! 3.48xl015 + Ycs-137/ 6.0xl017 ::S 1 Where Yco-60 is the source strength of Co-60 and Ycs-137 is the source strength of Cs-137 NUH09.0101 A.7-18b All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 Table A.7-3 ________ --- ----Appendices Containing LoadfugErocedures-for-\\larious-DSCs-DSC Model Annendix NUHOMS-24PT4 A.7.7.1 NUHOMS-32PT A.7.7.2 NUHOMS-24PTH A.7.7.3 NUHOMS-32PTH A.7.7.4 NUHOMS-32PTH1 A.7.7.5 NUHOMS-37PTH A.7.7.6 NUHOMS-61BT A.7.7.7 NUHOMS-61BTH A.7.7.8 NUHOMS-69BTH A.7.7.9 RWC A.7.7.10 Table A.7-4 Appendices Containing Unloading Procedures for Various DSCs DSC Model Annendix NUHOMS-24PT4 A.7.7.1, Section A.7.7.1.4 NUHOMS-32PT A.7.7.2, Section A.7.7.2.4 NUHOMS-24PTH A.7.7.3, Section A.7.7.3.4 NUHOMS-32PTH A.7.7.4, Section A.7.7.4.4 NUHOMS-32PTH1 A.7.7.5, Section A.7.7.5.4 NUHOMS-37PTH A.7.7.6, Section A.7.7.6.4 NUHOMS-61BT A.7.7.7, Section A.7.7.7.4 NUHOMS-61BTH A.7.7.8, Section A.7.7.8.4 NUHOMS-69B1H A.7.7.9, Section A.7.7.9.4 NUH09.0101 A.7-18c

MP197 Transportation Packaging Safety Analysis Report Rev. 18D, 12/18 I A.8.1.5.2 Gaskets The lid and all the other containment penetrations are sealed using 0-ring seals. Leakage testing of the seals is described in Section A.8.1.4.1. A.8.1.5.3 Impact Limiter Leakage Test Prior to initial use, the following test will be performed, after all the seal welds have been completed on the impact limiter to verify that the impact limiter wood is completely enclosed, thereby preventing any moisture exchange with the ambient environment. Each impact limiter container is pressurized to a pressure between 2.0 and 3.0 psig. Test all the weld seams and penetrations for leakage using a soap bubble test. A.8.1.5.4 Functional Tests The following functional tests will be performed prior to the first use of the cask. Generally these tests will be performed at the fabrication facility.

a. Installation and removal of the lid, ram access cover plates, port plugs, and other fittings will be observed. Each component will be checked for difficulties in installation and removal.

After removal, each component will be visually examined for damage. Any defects will be corrected prior to the acceptance of the cask.

b. After installation of the fuel basket into the DSC, each basket compartment will be checked by gauge to demonstrate that the fuel assemblies will fit in the basket.

A.8.1.6 Shielding Tests Chapter A.5 presents the analyses performed to ensure that the NUHOMS-MP197HB package shielding integrity is adequate. A.8.1.6.1 Gamma Shield Test i;=a scannin~-shall be used t~demonstrate the soundness of the gamma shielding that is

• poured into the annulus formed by the inner containment vessel and the outer shell of the 1 NUHOMS-MP 197HB prior to installation of the neutron shield.

l The test procedure provides description of the measuring technique, source type and strength used to measure the shield effectiveness, standards and methods used to calibrated the source, sensors and other pertinent equipment, grid pattern used to check the shield, type of gamma sensor used to measure shield effectiveness, specific test requirements and measurements, and acceptance criteria. l The acceptance criteria for the gamma scan shall be based on the results of measurements of a I calibration test block constructed to represent the MP 197HB cask gamma shield configuration ~ T~; calibration test block shall consists of the steel inner shell, lead, and steel outer shell. The ~bration test block is fabricated using nominal steel shell thicknesses and nominal lead NUH09.0101 A.8-4 All Indicated Changes are discussed in Enclosure 3, Item 2

MP197 Transportation Packaging Safety Analysis Report Rev.18D, 12/18 I thickness less 5% of the nominal thickness as provided in SAR drawings given in Chapter A. I,

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----- _______ --rAppendix A.-1.4..10.- The-calibr.ation testblock lead.thickness.shalLbe-2.85-inches.with-the. __ _

exception of MP l 97HB Unit O 1 for which the lead thickness shall be 2. 77 inches. The source and detector geometry used for the MP 197HB gamma shield test shall be the same as that used for the calibration test block. The measured counts for the MP l 97HB gamma shield shall be less than or equal!o the measured counts from the gamma scan of the calibration test block. I NUH09.0101 A.8-4a All Indicated Changes are discussed in Enclosure 3, Item 2 I ! -- *-* *- - -..

MP197 Transportation Packaging Safety Analysis Report Rev. ~8D, 12/18 I _I ___ ***-- --* A.8.1.6.2 Neutron Shield The radial neutron shield is protected from damage or loss by the aluminum and steel enclosure. The neutron shield material, VY AL B, is a proprietary vinyl ester resin mixed with alumina hydrate and zinc borate which are added for their fire retardant properties. The primary function of the resin is to shield against neutrons, which is performed primarily by the hydrogen content in the resin. The sole function of the boron is to suppress n-y reactions with hydrogen. The resin also provides some gamma shielding, which is a function of the overall resin density, and is not sensitive to composition. The proprietary process for the VY AL-B mixing and installation is described in SAR Section A.5.5. The following are acceptance values for density and chemical composition for the resin. The values used in the shielding calculations of Chapter A.5 are included for comparison. Chapter A.5 values Acceptance Testing Values Acceptance range Element Nominal wt% Element Wt% (wt%) H 4.54 H 5.0 +/-8 B 0.82 B 0.9 +/- 10 The minimum resin density in acceptance testing is 1.75 g/cm3* Resin composition or density test results which fall outside of this range will be evaluated to ensure that the shielding regulatory dose limits are not exceeded. Tests are performed at loading to ensure that the radiation dose limits are not exceeded for each cask. A.8.1.7 Neutron Absorber Tests The neutron absorber used for criticality control in the DSC baskets may consist of any of the following types of material. Depending on the DSC model, these neutron absorber materials may be used alone or be paired with aluminum: (a) Boron-aluminum alloy (borated aluminum) (b) Boron carbide/Aluminum metal matrix composite (MMC) (c) Boral These materials only serve as neutron absorber for criticality control and as heat conduction paths. The J'vIP197HB packaging safety analyses do not rely upon their mechanical strength. The radiation and temperature environment in the cask is not sufficiently severe to damage these metallic/ceramic materials. To assure performance of the neutron absorber's design function only NUH09.0101 A.8-5 All Indicated Changes are discussed in Enclosure 3, Item 2}}