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NUH32PHB-0101, Revision 4, Design Criteria Document (DCD) for the Nuhoms 32PHB System for Storage
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Site:	Calvert Cliffs
Issue date:	03/10/2015
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ENCLOSURE9 NUH32PHB-0101, Revision 4, Design Criteria Document (DCD) for the NUHOMS 32PHB System for Storage Calvert Cliffs Nuclear Power Plant March 10, 2015

CONTROLLED COPY E-281 A

TRANSNUCLEAR AN AREVA COMPANY DESIGN CRITERIA DOCUMENT PAGE: 1 of 67 DOCUMENT NO: NUH32PHB.0101 PROJECT NAME: High Bum-up NUHOMS32PHB System for PWR Fuel PROJECT NO: 10955 [CLIENT: Calvert Cliffs Nuclear Power Plant Inc.

(CCNPP)

DO.UMENT TITLE:

Design Criteria Document (DCD) for the NUHOMS 32PHB System for Storage

SUMMARY

DESCRIPTION:

This document specifies design requirements for the NUHOMS 32PHB system. The system consists of the Dry Shielded Canister (DSC), the Horizontal Storage Module, Model HSM-HB and the modified Calvert Cliffs Nuclear Power Plant Onsite Transfer Cask (CCNPP-FC TC).

This DCD supports the request of CCNPP to design an Independent Spent Fuel Storage Installation (ISFSI) to allow dry storage of high burnup fuel assemblies. This DCD presents the criteria for the NUHOMS 32PHB DSC, HSM-HB and the CCNPP-FC Transfer Cask.

PREPARER VERIFIER APPROVER REV. SIGNATUREIDATE SIGNATUREIDATE SIGNATUREIDATE 0 Kamran Tavassoli Prakash Narayanan Peter Shih I Kamran Tavassoli Prakash Narayanan Peter Shih 2 Kamran Tavassoli Raheel Haroon Peter Shih 3

Girish Patel Venkata Venigalla Peter Sh*i 4 62 -20 Girish Patel IS- Liu Hui Liu 7Peter Shi Shih /(

f

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REISION: 4 PROJECT NO: 10955 PAGE: 2 of 67 REVISION

SUMMARY

REV. DESCRIPTION AFFECTED PAGES 0 Initial Issue All I Revision I corrects the headings in Table 4-4 and clarifies the source terms for 1-2, 11, 13, 16, 62 reconstituted fuel assemblies. See DCR NUH32PHB-003, Rev. 0.

Revision 2 corrects the material of cladding for AREVA fuel assembly in Table 4-1 2 from Zircaloy to M5 and adds a reference for the 75g loads considered for side and 1, 2, 14, 38, 50 end drop accident conditions. See DCR NUH32PHB-009, Rev. 0.

3 Revision 3 removes incomplete statement from Section 14.0. 1, 2, 65 4 Revision 4 corrects temperature designation or symbol 1,2,42,58

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 3 of 67 TABLE OF CONTENTS Page 1.0 SCO PE .............................................................................................................................................................. 6 2.0 A PPLICA BLE D O CUM EN TS ............................................................................................................... 6 2.1 Codes and Standards ............................................................................................................................... 6 2.2 Federal, Regulations ................................................................................................................................ 7 2.3 N RC B ulletins, Regulatory G uides, N UREG D ocum ents ................................................................. 7 2.4 Technical Reports and Docum ents .................................................................................................... 9 3.0 G EN ERA L SY STEM DESC RIPTIO N .................................................................................................... 12 4.0 D ESIGN BA SIS FU EL CH A RA CTER ISTICS ......................................................................................... 13 5.0 G ENERA L D ESIGN REQ U IREM EN TS ................................................................................................ 20 5.1 System Design Features and Considerations .................................................................................. 20 5.1.1 Canister Features and Considerations ................................................................................ 20 5.1.2 H SM -H B Features and Considerations ............................................................................... 23 5.1.3 CCN PP-FC TC Features and Considerations ...................................................................... 25 6.0 EN VIRO N M EN TA L CON D ITION S ...................................................................................................... 26 6.1 D ead Load .............................................................................................................................................. 26 6.2 W ind, Tornado, and Snow .................................................................................................................... 26 6.3 Seismic .................................................................................................................................................. 27 6.4 Flood ..................................................................................................................................................... 27 6.5 Fire A ccident ........................................................................................................................................ 27 6.6 Forest Fire ............................................................................................................................................. 27 6.7 Therm al Environm ental Conditions ............................................................................................... 27 6.7.1 Fuel Handling B uilding Conditions .................................................................................... 27 6.7.2 ISFSI Site Conditions ........................................................................................................ 28 6.8 Other N atural Phenomena ..................................................................................................................... 28 7.0 O PERA TION AL CON D ITION S ................................................................................................................... 29 7.1 H SM -H B ............................................................................................................................................... 29 7.1.1 Normal and Off-Normal Operational Handling Loads ..................................................... 29 7.2 32PH B Canister .................................................................................................................................... 29 7.2.1 N orm al O perational Loads .................................................................................................. 29 7.2.2 O ff-N orm al Operational Loads ........................................................................................... 29 7.2.3 Accident O perational Loads ............................................................................................... 30 7.3 CCN PP-FC Transfer Cask .................................................................................................................... 30 7.4 Load Com binations ............................................................................................................................... 30 8.0 D SC STRU CTU RA L D ESIG N REQ UIREM EN TS .................................................................................. 34 8.1 N UH O MS32PH B D SC Structural D esign Criteria ...................................................................... 34 8.1.1 N U H OM S 32PH B D SC Shell Stress Lim its ................................................................... 34 8.1.2 N UH OM S32PH B Canister Basket Stress Lim its ............................................................ 34 8.2 Fuel Assem bly Evaluations .................................................................................................................. 36 8.3 Weld stresses ........................................................................................................................................ 36

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NH32PHB.0101 REVISION: 4 PROJECT NO: 10955l PAGE: 4 of 67 8.4 Canister Loads Descriptions ........................................................................................................... 36 8.4.1 Deadweight .............................................................................................................................. 36 8.4.2 Internal Pressure ...................................................................................................................... 37 8.4.3 Therm al .................................................................................................................................... 37 8.4.4 Seism ic .................................................................................................................................... 37 8.4.5 Handling .................................................................................................................................. 37 8.4.6 Drop Loads .............................................................................................................................. 38 8.4.7 Flood Loads ............................................................................................................................. 38 9.0 ONSITE TRANSFER CASK STRUCTURAL DESIGN REQUIREMENTS ......................................... 49 9.1 Structural Design Criteria ..................................................................................................................... 49 9.2 Loads and Load Combinations ...................................................................................................... 49 9.2.1 Deadweight .............................................................................................................................. 49 9.2.2 Internal Pressure ...................................................................................................................... 49 9.2.3 Therm al .................................................................................................................................... 49 9.2.4 Seism ic .................................................................................................................................... 50 9.2.5 Handling .................................................................................................................................. 50 9.2.6 Drop Loads .............................................................................................................................. 50 9.2.7 Tornado W ind and M issile Loads ...................... :..................................................................... 50 9.2.8 Flood Loads ............................................................................................................................. 50 10.0 HSM -HB STRUCTURA L DESIGN CRITERIA ..................................................................................... 60 11.0 THERM AL REQUIREM ENTS ..................................................................................................................... 60 12.0 SHIELDIN G REQUIREM ENTS ................................................................................................................... 62 13.0 CRITICALITY REQUIREM ENTS ......................................................................................................... 63 13.1 General Criticality Criteria ................................................................................................................... 63 14.0 CON FINEM ENT/CON TAIN M ENT CRITERIA .................................................................................... 65 15.0 ACCEPTAN CE TESTIN G ............................................................................................................................ 65 16.0 M ATERIAL REQUIREM ENTS .................................................................................................................... 66 16.1 Specifications ........................................................................................................................................ 66 16.2 Properties .............................................................................................................................................. 66 16.3 Impact Properties Test .......................................................................................................................... 66 16.4 Materials Suitability (Chem ical, Galvanic and Other Reactions) ................................................... 66 16.5 Protective Coatings ............................................................................................................................... 67 16.6 Emissivities ........................................................................................................................................... 67 16.7 Effects of Radiation .............................................................................................................................. 67 16.8 Prohibited / Hazardous M aterials .................................................................................................... 67 17.0 QU ALITY ASSURANCE REQ UIREM EN TS ......................................................................................... 67

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 TREVISION: 4 PROJECT NO: 10955 PAGE: 5 of 67 LIST OF TABLES Page Table 4-1 Fuel Assembly Design Characteristics ............................................................................................. 14 Table 4-2 Fuel Types C haracteristics ..................................................................................................................... 14 Table 4-3 Fuel Assem bly Region Lengths ....................................................................................................... 15 Table 4-4 Bounding Neutron Sources per Fuel Assembly for 1000 Watt and 800 Watt .................................. 16

'Fable 4-5 Bounding Gamma Source Terms (Gamma/Sec) per Fuel Assembly for 800 Watt .......................... 17 Table 4-6 Bounding Gamma Source Terms (Gamma/Sec) per Fuel Assembly for 1,000 Watt ...................... 18 Table 7-1 DSC Structural Loading Conditions .................................................................................................. 31 Table 7-2 Summary of 32PHB DSC Shell Load Combinations ....................................................................... 32 Table 7-3 Summary of 32PHB Basket Load Combinations ............................................................................ 33 Table 8-1 Material Properties - SA-240/SA-479 Type 304 (DSC) ................................................................... 39 Table 8-2 Material Properties - Aluminum 6061 (DSC) ................................................................................. 40 Table 8-3 Analysis Properties for Aluminum Transition Rails (DSC) [2.4.26] ................................................ 40 Table 8-4 Material Properties - Aluminum 1100 (DSC) ................................................................................. 41 Table 8-5 Material Properties - Helium (DSC) ............................................................................................... 42 Table 8-6 M aterial Properties - N itrogen ........................................................................................................ 42 Table 8-7 Summary of Stress Criteria for Subsection NB Pressure Boundary Components ............................. 43 Table 8-8 Summary of Stress Criteria for Subsection NG Components (Austenitic) ....................................... 45 Table 8-9 32PH B DSC Pressure Loads .................................................................................................................. 46 Table 8-10 Thermal Conditions for 32PHB DSC Analyses ............................................................................. 47 Table 8-11 H andling Loads .................................................................................................................................... 48 Table 9-1 Material Properties - SA-240/SA-479 Type 304 ASTM A-240, Type 304 (TC) ............................ 51 Table 9-2 Material Properties - SA 516, Gr. 70 (TC) ...................................................................................... 52 Table 9-3 Material Properties - SA 564, Gr. 630 (TC) .................................................................................... 53 Table 9-4 Material Properties - SA-182 Type F304N (TC) ............................................................................ 54 Table 9-5 Material Properties - SA-193 Gr. B7 (TC) ...................................................................................... 55 Table 9-6 Mechanical Properties for ASTM B29 Lead (DSC and TC) ............................................................ 56 Table 9-7 Thermal Properties for Lead (DSC and TC) .................................................................................... 57 Table 9-8 Mechanical and Thermal Properties for NS-3 (TC) .......................................................................... 57 Table 9-9 Material Properties - Air (DSC and TC) ........................................................................................... 58 Table 9-10 Structural Stress Criteria for Transfer Cask .................................................................................... 59 Table 9-11 Structural Stress Criteria for Transfer Cask Bolts .......................................................................... 59 LIST OF FIGURES Page Figure 4-1 Heat Load Zone Configuration for the Maximum Heat Load ......................................... 19

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 EISION:

ý _4 PROJECT NO: 10955] PAGE: 6 of 67 1.0 SCOPE This Design Criteria Document (DCD) specifies the design requirements of the NUHOMS' 32PHB Dry Shielded Canister (DSC) for storage, the HSM-HB Horizontal Storage Module, and the CCNPP-FC on-site transfer cask (TC). The system is designed for high bumup fuel, up to 62 GWD/MTU, with a maximum assembly average initial enrichment of 5% wt U-235.

General design requirements include structural, thermal, nuclear criticality safety, confinement/containment, and radiological protection criteria.

2.0 APPLICABLE DOCUMENTS 2.1 Codes and Standards 2.1.1 ASME Boiler and Pressure Vessel Code,Section II, "Materials Specifications," Parts A, B, C and D, 1998 edition with all addenda up to and including 1999 Addenda.

2.1.2 ASME Boiler and Pressure Vessel Code, Section 111, Division 1, Subsections NB, NG, and NC, 1998 edition with all addenda up to and including 1999 Addenda.

2.1.3 ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsections NC and Appendices,Section II, Part D, 1992 edition.

2.1.4 ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 1998 edition with all addenda up to and including 1999 Addenda.

2.1.5 ASME Boiler and Pressure Vessel Code,Section V, "Nondestructive Examination," 1998 edition with all addenda up to and including 1999 Addenda.

2.1.6 ASME Boiler and Pressure Vessel Code,Section IX, "Welding and Brazing Qualifications," 1998 edition with all addenda up to and including 1999 Addenda.

2.1.7 ANSI Y14.5M, "Dimensions and Tolerancing," 1982, 2.1.8 ANSI N14.5, "Leakage Tests on Packages for Shipment of Radioactive Materials," 1997.

2.1.9 ANSI N14.6, "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More", 1978 (and 1993).

2.1 .10 ANSI 8.17, "Criticality Safety Criteria for Handling, Storage, and Transportation of LWR Fuel Outside reactors," 1984.

2.1 .11 ANSI N 16.9, "Validation of Calculational Methods for Nuclear Criticality Safety."

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PB.0101 TREVISION: 4 PROJECT NO: 10955 1PAGE: 7 of 67 2.1.12 American National Standards Institute, American Nuclear Society, ANSI/ANS 57.9 - 1992, "Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type)".

2.1.13 American Concrete Institute, ACI 349 - 97, "Code Requirements for Nuclear Safety Related Concrete Structures."

2.1.14 American Institute of Steel Construction, "AISC Manual of Steel Construction," Ninth Edition.

2.1.15 American Society of Civil Engineers, ASCE 7-95, "Minimum Design Loads for Buildings and Other Structures," (formerly ANSI A58. 1).

2.1.16 American Welding Society, AWS D 1.6 - 1999, "Structural Welding Code - Stainless Steel."

2.1.17 American Welding Society, AWS D1.I - 1988, "Structural Welding Code - Steel."

2.1.18 American Welding Society, AWS A2.4- 1986, "Weld Symbols."

2.2 Federal Regulations 2.2.1 Title 10, Code of Federal Regulations, Part 71, "Packaging and Transportation of Radioactive Materials."

2.2.2 Title 10, Code of Federal Regulations, Part 72, "Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation."

2.2.3 Title 10, Code of Federal Regulations, Part 20, "Standards for Protection Against Radiation."

2.2.4 Title 10, Code of Federal Regulations, Part 50, "Domestic Licensing of Production and Utilization Facilities."

2.3 NRC Bulletins, Regulatory Guides, NUREG Documents

Note - NUREG documents are for guidance only, these documents do not impose requirements.

2.3.1 NRC Bulletin 96-04, "Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks," July 5, 1996.

2.3.2 NRC Regulatory Guide 1.13, "Spent Fuel Facility Design Basis."

2.3.3 NRC Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants."

2.3.4 NRC Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Power Plants."

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NtH32PHB.0101 7 REVISION: 4 PROJECT NO: 10955 1 PAGE: 8 of 67 2.3.5 NRC Regulatory Guide 1.76, "Design Basis Tornado for Nuclear Power Plants."

2.3.6 NRC Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response Analysis."

2.3.7 NRC Regulatory Guide 1.145, "Atmospheric Dispersement Models for Potential Accident Consequence Assessments at Nuclear Power Plant," February 1989.

2.3.8 NRC Regulatory Guide 3.48, "Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage)."

2.3.9 NRC Regulatory Guide 3.53, "Applicability of Existing Regulatory Guides to the Design and Operation of an Independent Spent Fuel Storage Installation."

2.3.10 NRC Regulatory Guide 3.60, "Design of an Independent Spent Fuel Storage Installation (Dry Storage Type)."

2.3.11 NRC Regulatory Guide 7.4, "Leakage Tests on Packages for Shipments of Radioactive Materials."

2.3.12 NRC Regulatory Guide 7.6, "Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels."

2.3.13 NRC Regulatory Guide 7.8, "Load Combinations for the Structural Analysis of Shipping Casks."

2.3.14 NRC Regulatory Guide 7.9, "Standard Format and Content of Part 71 Applications for Approval of Packaging for Radioactive Material."

2.3.15 Regulatory Guide 7.11, "Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 inches (0.1m)."

2.3.16 NRC Regulatory Guide 7.12, "Fracture Toughness Criteria of Base Material for Steel Shipping Cask Containment Vessels."

2.3.17 NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," July 1980.

2.3.18 NUREG-0800, Standard Review Plan, Section 3.3.1 "Wind Loading" and Section 3.5.1.4 Missiles Generated by Natural Phenomenon."

2.3.19 NUREG-1536, "Standard Review Plan for Dry Cask Storage Systems - Final Report," U.S.

Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, January 1997.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 EISION:

ý 4_

PROJECT NO: 10955[ PAGE: 9 of 67 2.3.20 NUREG-1567, "Standard Review Plan for Spent Fuel Dry Storage Facilities - Final Report," U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, March 2000.

2.3.21 NUREG/CR-5661, "Recommendations for Preparing the Criticality Safety Evaluation of Transport Packages", April 1997.

2.3.22 NUREG/CR-6487, "Containment Analysis for Type B Packages Used To Transport Various Contents", November 1996.

2.3.23 Interim Staff Guidances (ISGs).

2.3.24 NUREG/CR-0481, SAND 77-1872, "An Assessment of Stress-Strain Data Suitable for Finite Element Elastic-Plastic Analysis of Shipping Casks," Sandia National Laboratories, September 1978.

2.3.25 NUREG/CR-2018, SAND 80-1870, "A Comparison of Analytical Techniques for Analyzing a Nuclear Spent-Fuel Shipping Cask Subjected to an End-On Impact," Sandia National Laboratories, June 1981.

2.3.26 NUREG/CR-3854, UCRL-53544, "Fabrication Criteria for Shipping Containers,"

Lawrence Livermore National Laboratories, March 1985.

2.3.27 NUREG/CR-3966, UCID-20639, "Methods for Impact Analysis of Shipping Containers,"

Lawrence Livermore National Laboratories, November 1987.

2.3.28 NUREG/CR-6407, "Quality Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety."

2.3.29 NUREG-7665 10, SAND76-0427, "Shock and Vibration Environments for Large Shipping Containers on Rail Cars and Trucks", Sandia National Laboratories, June 1977.

2.3.30 NUREG/CR-6322, "Buckling Analysis of Spent Fuel Basket", LLNL, May, 1995.

2.4 Technical Reports and Documents 2.4.1 NUTECH Report NUH-002, "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS - 24," Revision I A, July 1989.

2.4.2 CCNPP Specification SP-0564, "Spent Fuel Storage Capacity Design Specification,"

Revision 10.

2.4.3 CCNPP Specification SP-0564C, "NUHOMS-32P Dry Shielded Canister," Revision 4.

2.4.4 CCNPP Specification SP-0564D, "Design Specification for NUHOMS 32PHB DSC (High Bum-up Dry Shielded Canister)," Revision 0.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955I PAGE: 10 of 67 2.4.5 Calvert Cliffs Nuclear Power Plant ISFSI, "Updated Safety Analysis Report," Docket Numbers 50-317 and 50-318, Revision 17.

2.4.6 Calvert Cliffs Nuclear Power Plant ISFSI, "Appendix A to Material License SNM-2505 Technical Specification," Amendment 7, NRC Docket No. 72-8.

2.4.7 Constellation Energy, Nuclear Generation Group, Nuclear Ahalysis Unit, Fleet Nuclear Fuels, Memorandum, "ISFSI 32PHB Source Term Input Data," Constellation Tracking No.

DE10290, TN Document No. NUH32PHB-0 102.

2.4.8 Constellation Energy, Nuclear Generation Group, Nuclear Analysis Unit, Fleet Nuclear Fuels, Attachment to the Email from Eric Yin to Sue Buyaskas, Dated January 27, 2010, TN Document No. NUH32PHB-0 105.

2.4.9 Constellation Energy, Nuclear Generation Group, Nuclear Analysis Unit, Fleet Nuclear Fuels, Email from John Massari to Sue Buyaskas, Dated March 16, 2010, TN Document No. NUH32PHB-0106.

2.4.16 Electric Power Research Institute Report NP-7419 Project 2813-9, "Fuel Assembly Behavior Under Dynamic Impact Loads due to Dry Storage Cask Mishandling," Final Report, July 1991.

2.4.11 CCNPP Letter "DES Support for Increased Control Element Assembly (CEA) Weight,"

dated March 27th, 2001, NEU 01-047.

2.4.12 Transnuclear, Inc., "Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," NRC Docket No. 72-1004, Transnuclear Document No. NUH-003, Revision 11.

2.4.13 Transnuclear Inc., "Updated Final Safety Analysis Report for the NUHOMS HD Horizontal Modular Storage System for Irradiated Nuclear Fuel", NRC Docket No. 72-1030, Revision 1.

2.4.14 Transnuclear Inc., "Design Criteria Specification (DCS) for the CoC 72-1004 (OS 197 and OS200) On-Site Transfer Casks," Transnuclear Specification No. NUH06-01 10, Revision 1.

2.4.15 Transnuclear Inc., "Design Criteria Specification (DCS) for the NUHOMS 32PTHI for Transportation and Storage," Transnuclear Specification No. NUH32PTH 1-0101, Revision 0.

2.4.16 Nuclear Regulatory Commission, "Safety Evaluation Report of Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," USNRC Docket Number 72-1004, Amendment No. 10.

2.4.17 Transnuclear, Inc., "NUHOMS System Operations Manual," NUH-07-118, Revision 5.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 11 of 67 2.4.18 Transnuclear, Inc., "Fabrication Specification for NUHOMS System Dry Shielded Canister," TN Specification NUH-03-105, Revision 10.

2.4.19 Transnuclear, Inc., "Precast Concrete Construction for NUHOMS HSM," TN Specification NUH-03-0214, Revision 3.

2.4.20 Schwartz, M. W., Witte, M. C., "Spent Fuel Cladding Integrity During Dry Storage",

UCID-21181, Lawrence Livermore National Laboratory, September 1987.

2.4.21 Transnuclear West (TNW) Inc., Document No. 3 1-B9604.97-003, dated December 19, 1997; Addendum to TNW Document No. 31 -B9604.0102, Rev. 2, An Assessment of Chemical, Galvanic and Other Reactions in NUHOMS Spent Fuel Storage and Transportation Casks.

2.4.22 Nuclear Assurance Corporation, "Domestic Light Water Reactor Fuel Design Evolution,"

Volume III, September 1981.

2.4.23 Adkins, H. E. Jr., et al, "Spent Nuclear Fuel Structural Response When Subject to an End Impact Accident," PVP2004-2804, PVP-Vol. 483, Transportation, Storage, and Disposal of Radioactive Materials-2004, July 25-29, 2004, San Dieg6 California USA 2.4.24 Geelhood, K. J. and Beyer, C. E., "PNNL Stress/Strain Correlation for Zircaloy," Pacific Northwest National Laboratory, March 2005.

2.4.25 Lawrence Livermore National Laboratory, "Dynamic Impact Effects on Spent Fuel Assemblies," LLNL/UCID Report No. 21246, October 1987.

2.4.26 Kaufman, J.G., ed., "Properties of Aluminum Alloys: Tensile, Creep, and Fatigue Data and High and Low Temperatures", The Aluminum Association (Washington, D.C.) and ASM International (Metals Park, Ohio), 1999.

2.4.27 ASME NQA-1, "Quality Assurance Requirements for Nuclear Facility Applications".

2.4.28 Welding Research Council (WRC), "Local Stresses in Spherical and Cylindrical Shells Due To External Loadings", Bulletin 107 August 1965, March 1979 Revision.

2.4.29 Tietz, T. E., "Determination of the Mechanical Properties of a High Purity Lead and a 0.058 Percent% Copper-Lead Alloy," Presented at the Sixty Second Annual Meeting of the ASTM Society, June 1959, ASTM 59, 1052.WADC Technical Report 57-695, ASTIA Document No. 151165, Stanford Research Institute, Menlo Park, CA, April, 1958.

2.4.30 Constellation Energy, Nuclear Generation Group, Calculation, "Comparison of the Radiological, Thermal, and Reactivity Characteristics of Assemblies with Missing or Inert Fuel Rods with the 32P ISFSI DSC Design Basis," CCNPP, Calculation No. CA06367, Rev. 0000, TN Document No. NUH32PI1B-0109.

2.4.31 Constellation Energy, Nuclear Generation Group, CCNPP Site, ECP Document, ES200100208, Rev. 0000, TN Document No. NUH32PHB-0 108.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 12 of 67 3.0 GENERAL SYSTEM DESCRIPTION The NUHOMS 32PHB system consists of the 32PHB DSC, the HSM-HB and the CCNPP-FC TC.

The 32PHB system is designed to allow transfer of spent fuel in the 32PHB DSC using the CCNPP-FC TC, provide for storage of spent fuel in an HSM-HB. The system will be licensed for storage in accordance with the requirements of Title 10, Part 72 (10 CFR'72) of the Code of Federal Regulations, via licensing amendment to the CCNPP ISFSI Final Safety Analysis Report [2.4.5].

The 32PHB DSC will accommodate up to 32 intact CE 14x14 or equivalent reload spent fuel assemblies (including fuel with stainless steel replacement rods) with and without axial blankets.

The NUHOMS 32PHB DSC design is similar to the NUHOMS 32P DSC design documented in

[2.4.5] with the maximum decay heat load per canister increased from 21.12 kW to 29.6 kW. The DSC has a nominal outside diameter and length (including grapple ring) of 67.25 inches and 176.5 inches, respectively. Solid aluminum transition rails are incorporated into the 32PHB basket to accommodate heat loads up to 29.6 kW.

The NUHOMS32PHB DSC consists of a shell assembly, which provides confinement and shielding, and an internal basket assembly, which locates and supports the fuel assemblies, transfers the heat to the cask body wall, and provides for criticality control as necessary to satisfy nuclear criticality safety requirements. The basket is a tube assembly, with aluminum and poison plates in between the tubes for heat transfer and criticality control. Except for the solid aluminum rails, the 32PHB basket is identical to the 32P basket documented in [2.4.5].

The HSM-HB to be used for the 32PHB system is similar to the horizontal storage module HSM-H with flat stainless steel heat shields described in the UFSAR for standardized NUHOMS System

[2.4.12], Appendix P and the UFSAR for NUHOMS HD System [2.4.13]. In these systems the HSM-H is used to store a 24PTH DSC (with a maximum canister length of 192.55" and canister diameter of 67.19"), or a 32PTH DSC (with a maximum canister length of 185.75" and canister diameter of 69.75"). As noted above the maximum length of a 32PHB DSC (including the grapple ring) is 176.5". The HSM-H internal cavity design has the flexibility to accommodate a shorter canister length with minor changes to the design of the rail spacer at the back end of the steel support structure. The HSM-HB with these modifications and flat stainless steel heat shields shall be evaluated as part of the 32PHB system analysis.

The transfer cask to be used for the 32PHB is the CCNPP-FC TC. The nominal cavity inner diameter and inner cavity length of CCNPP-FC TC are 68.0" and 173.5", respectively. These dimensions are identical to the corresponding dimensions of the existing CCNPP TC. The cask lid of CCNPP-FC TC is redesigned to improve the cask's thermal performance. The new lid contains small openings around the periphery that vent out forced air that is injected at the bottom of the cask (through the ram access opening) and circulates up through the cask's length through the cask/DSC annulus. A 0.5 inch thick spacer disc with wedge shaped protrusions is installed at the bottom of the TC to facilitate air flow coming through the ram access opening to the annular space around the DSC.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0 10 1 REVISION: 4

PROJECT NO: 10955[ PAGE: 13 of 67 4.0 DESIGN BASIS FUEL CHARACTERISTICS The NUHOMS 32PHB DSC shall be designed for PWR fuel assemblies with characteristics as described in Table 4-1 and Table 4-2.

The 32PHB DSC is designed to accommodate up to 32 intact PWR fuel assemblies. The DSC payload may not include non-fuel assembly hardware'components such as Control Element Assemblies (CEAs).

The bounding radiological source terms for fuel assemblies are shown in Table 4-4, Table 4-5, and Table 4-6. Based on evaluations performed in [2.4.30] and [2.4.31 ], the source terms presented in these tables are bounding for reconstituted fuel assemblies provided that the restriction on the number of reconstituted fuel rods and requirements for additional cooling time specified in [2.4.30]

and [2.4.3 1] are followed.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 14 of 67 Table 4-1 Fuel Assembly Design Characteristics Assembly Class Standard Value Added Pellet AREVA Assembly ______Class___ CE 14x04 (VAP) AREVA Clad Material [2.4.7] Zircaloy-4 Zircaloy-4 M5 Pellet stack U0 2 density (%TD) [2.4.7] 93.5 - 96% 96% 96%

Number of Rods 176 176 176 Number of Water Holes 5 5 5 Fuel rod pitch (in) [2.4.71 0.580 0.580 0.580 Pellet OD (in) [2.4.7] 0.3765 0.381 0.3805 Clad ID (in) [2.4.7] 0.384 0.388 0.387 Clad OD (in) [2.4.7] 0.440 0.440 0.440 Guide tube ID (in) [2.4.71 1.035 1.035 1.035 Guide tube OD (in) [2.4.7] 1.115 1.115 1.115 Maximum Enrichment (wt% U-235) [2.4.7] 4.50% 5% 5%

Table 4-2 Fuel Types Characteristics Description Value Physical Characteristics Maximum Assembly Weight 1375 lbs Maximum Assembly Length (including 158 inches irradiation growth)

Number of Grid Spacers 9 Nominal Assembly Envelope 8.25 inches square Active Fuel Length 136.7 inches Radiological Source Maximum Peak Pin Burnuo 62 GWd/MTU Assembly Average Bumup 58 GWd/MTU Maximum initial enrichment 5.0 wt% U-235 Minimum initial enrichment 2.0 wt% U-235 Maximum Uranium Content 420 kg/assembly As needed to reach 0.8 to 1.0 kW per Assembly. See Figure 4-1.

Nominal Specific Power 32.2 MW/MTU Maximum Neutron Source per Assembly 6.66 x 108 n/sec Maximum In-core Gamma Source per 7.45 x 1015 y/sec Assemblv Thermal Source Maximum Heat Load per Assembly 0.8 to 1.0 kW depending on Zone, See Figure 4-1

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: _4 PROJECT NO: 10955l PAGE: 15 of 67 Table 4-3 Fuel Assembly Region Lengths Unit inch ")

FuelType Fuel Type Fuel Type Fuel Type Fuel Type AREVA Standard Guardian VAP(Guardian)

Region Region Length Region Length Region Length Region Length' Top Ending Fitting 6.469 5.766 5.491 6.483 Plenum 9.217 10.525 9.804 8.816 Active Fuel 136.7 136.7 136.7 136.7 Bottom Ending Fitting 4.486 4.25 5.246 4.873 Sum 156.872 157.241 157.241 156.872 Notes:

fl) The data in Table 4-3 is provided by Constellation Energy, Nuclear Generation Group in [2.4.81.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVIS:ION: _4 PROJECT NO: 10955 PAGE: 16 of 67 Table 4-4 Bounding Neutron Sources per Fuel Assembly for 1000 Watt and 800 Watt 1000W 800W Adiusted for 800W Emin (MeV) Emax (MeV) Neutrons/Sec Neutrons/Sec Neutrons/Sec 1.40E+O1 2.OOE+01 O.00E+00 O.OOE+00 0.OOE+00 1.20E+01 1.40E+O 1 1.1 8E+05 9.OOE+04 9.46E+04 i.00E+01 1.20E+0I 7.20E+05 5.52E+05 5.80E+05 8.OOE+00 1.00E+01 2.44E+06 1.87E+06 1.97E+06 7.50E+00 8.OOE+00 1.97E+06 1.51E+06 1.59E+06 7.OOE+00 7.50E+00 2.64E+06 2.02E+06 2.12E+06 6.50E+00 7.OOE+00 3.90E+06 2.99E+06 3.14E+06 6.OOE+00 6.50E+00 5.83E+06 4.47E+06 4.70E+06 5.50E+00 6.OOE+00 8.74E+06 6.70E+06 7.04E+06 5.OOE+00 5.50E+00 1.18E+07 9.OOE+06 9.46E+06 4.50E+00 5.OOE+00 1.63E+07 1.25E+07 1.31E+07 4.OOE+00 4.50E+00 2.13E+07 1.63E+07 1.71E+07 3.50E+00 4.OOE+00 3.42E+07 2.63E+07 2.76E+07 3.OOEj0O 3.50E+00 4.24E+07 3.26E+07 3.43E+07 2.50E+00 3.OOE+00 5.5 1E+07 4.25E+07 4.47E+07 2.35E+00 2.50E+00 2.08E+07 1.61E+07 1.69E+07 2.15E+00 2.35E+00 2.92E+07 2.25E+07 2.36E+07 2.OOE+00 2.15E+00 2.32E+07 1.79E+07 1.88E+07 1.80E+00 2.OOE+00 3.39E+07 2.60E+07 2.74E+07 1.66E+00 1.80E+00 2.62E+07 2.01E+07 2.1 IE+07 1.57E+00 1.66E+00 1.74E+07 1.34E+07 1.41E+07 1.50E+00 1.57E+00 1.45E+07 1.12E+07 1.17E+07 1.44E+00 1.50E+00 1.24E+07 9.53E+06 1.00E+07 1.33E+00 1.44E+00 2.49E+07 1.91E+07 2.01E+07 1.20E+00 1.33E+00 3.08E+07 2.36E+07 2.48E+07 1.00E+00 1.20E+00 4.78E+07 3.67E+07 3.86E+07 8.OOE-01 1.00E+00 4.65E+07 3.57E+07 3.75E+07 7.OOE-0 I 8.OOE-01 2.69E+07 2.06E+07 2.17E+07 6.OOE-01 7.OOE-01 2.68E+07 2.06E+07 2.16E+07 5.12E-01 6.OOE-01 2.31E+07 1.77E+07 1.86E+07 5.1OE-01 5.12E-01 5.25E+05 4.02E+05 4.23E+05 4.50E-01 5.1 OE-0 I 1.57E+07 1.21 E+07 1.27E+07 4.OOE-01 4.50E-01 1.31E+07 1.01E+07 1.06E+07 3.OOE-0I 4.OOE-01 2.53E+07 1.94E+07 2.04E+07 2.OOE-01 3.OOE-01 5.23E+03 4.72E+03 4.96E+03 1.50E-01 2.OOE-01 2.61E+03 2.36E+03 2.48E+03 1.OOE-01 1.50E-01 2.61E+03 2.36E+03 2.48E+03 7.50E-02 1.OOE-01 0.OOE+00 O.OOE+00 0.OOE+00 7.OOE-02 7.50E-02 0.OOE+00 0.OOE+00 0.OOE+00 6.OOE-02 7.OOE-02 O.OOE+00 0.OOE+00 0.OOE+00 4.50E-02 6.OOE-02 0.OOE+00 0.OOE+00 0.OOE+00 3.OOE-02 4.50E-02 0.OOE+00 0.OOE+00 O.OOE+00

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH*32PHB.0101 TREVI1SION: 4 PROJECT NO: 10955 1PAGE: 17 of 67 2.OOE-02 3.OOE-02 O.OOE+00 O.OOE+00 O.OOE+00 1.00E-02 2.OOE-02 O.OOE+00 O.OOE+00 O.OOE+00 Total 6.66E+08 5.12E+08 5.38E+08 Notes:

) Neutron Source is provided by Constellation Energy, Nuclear Generation Group in [2.4.9].

Table 4-5 Bounding Gamma Source Terms (Gamma/Sec) per Fuel Assembly for 800 Watt Active Fuel LEF Plenum UEF Total Emin Emax Gamma/sec Gamma/sec Gamma/sec Gamma/sec (MeV) (MeV) Gamma/sec O.OOE+00 to 2.OOE-02 1.51E+15 9.76E+11 2.14E+11 4.75E+11 1.51E+15 2.OOE-02 to 3.OOE-02 3.33E+14 3.96E+12 1.19E+I I 3.88E+12 3.41E+14 3.OOE-02 to 4.50E-02 3.94E+14 7.77E+1l 3.79E+10 7.20E+1I 3.96E+14 4.50E-02 to 7.OOE-02 2.66E+14 7.49E+10 1.94E+10 2.64E+10 2.66E+14 7.OOE-02 to 1.00E-01 1.89E+14 3.58E+10 9.20E+09 1.27E+10 1.89E+14 l.00E-01 to 1.50E-01 2.08E+14 3.37E+10 4.77E+09 2.27E+10 2.08E+14 1.50E-01 to 3.OOE-01 1.69E+14 2.51E+I I 6.58E+09 2.48E+11 1.69E+14 3.OOE-0I to 4.50E-01 8.96E+13 1.47E+12 3.29E+ 10 1.47E+ 12 9.26E+13 4.50E-01 to 7.OOE-01 2.26E+15 1.89E+12 4.20E+10 1.89E+12 2.26E+15 7.OOE-0I to 1.00E+00 4.63E+14 8.46E+10 1.05E+I I 1.18E+10 4.63E+14 1.00E+00 to 1.50E+00 8.67E+13 2.08E+13 7.37E+12 1.85E+12 1.17E+14 1.50E+00 to 2.OOE+00 3.70E+12 8.59E+03 4.04E+03 1.91E+03 3.70E+12 2.OOE+00 to 2.50E+00 2.91E+12 1.1OE+08 3.89E+07 9.75E+06 2.91E+12 2.50E+00 to 3.OOE+00 7.19E+ 10 1.70E+05 6.04E+04 1.51E+04 7.19E+ 10 3.OOE+00 to 4.OOE+00 8.87E+09 5.23E-10 1.04E-13 1.32E-12 8.87E+09 4.OOE+00 to 6.OOE+00 3.48E+06 O.OOE+00 O.OOE+00 O.OOE+00 3.48E+06 6.OOE+00 to 8.OOE+00 4.OOE+05 O.OOE+00 O.OOE+00 O.OOE+00 4.OOE+05 8.OOE+00 to 1.1OE+O1 4.60E+04 O.OOE+00 O.OOE+00 O.OOE+00 4.60E+04 Total 5.97E+15 3.03E+13 7.97E+12 1.06E+13 6.02E+15 Notes:

(1) Gamma Source is provided by Constellation Energy, Nuclear Generation Group in [2.4.9].

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 ]REVISION: 4 PROJECT NO: 10955I PAGE: 18 of 67 I Table 4-6 Bounding Gamma Source Terms (Gamma/See) per Fuel Assembly for 1,000 Watt Active Fuel LEF Plenum UEF Total E(me E(Ma Gamma/sec Gamma/sec Gamma/sec Gamma/sec Gamma/see (M~eV) (MeV)

O.OOE*00 to 2.OOE-02 1.83E+15 1.20E+12 2.60E+ 11 5.96E+1I 1.83E+15 2.OOE-02 to 3.OOE-02 4.03E+14 5.01E+12 1.49E+: I 4.91E+12 4.13E+14 3.OOE-02 to 4.50E-02 4.78E+14 9.52E+ 11 4.59E+10 8.85E+I 1 4.80E+14 4.50E-02 to 7.OOE-02 3.22E+14 9.05E+10 2.33E+10 3.25E+10 3.23E+14 7.OOE-02 to 1.00E-0I 2.31E+14 4.32E+10 1.10E+10 1.56E+10 2.31E+14 I.OOE-01 to 1.50E-01 2.58E+14 4.11E+I0 5.74E+09 2.79E+10 2.58E+14 1.50E-01 to 3.OOE-0I 2.07E+14 3.08E+11 8.05E+09 3.05E+11 2.08E+14 3.OOE-01 to 4.50E-01 1.11E+14 1.81E+12 4.04E+10 1.81E+12 1.14E+14 4.5OE-01 to 7.OOE-01 2.79E+15 2.33E+12 5.16E+10 2.33E+12 2.79E+15 7.OOE-01 to 1.00E+00 6.42E+14 1.14E+ II 1.41E+I1 1.59E+10 6.42E+14 1.00E+00 to 1.50E+00 1.14E+14 2.49E+13 8.85E+12 2.21E+12 1.50E+14 1.50E+00 to 2.OOE+00 4.94E+12 1.97E+04 8.73E+03 5.22E+03 4.94E+12 2,OOE+00 to 2.50E+00 3.82E+12 1.31E+08 4,67E+07 1.17E+07 3.82E+12 2.50E+00 to 3.OOE+00 I.00E+11 2.03E+05 7.24E+04 1.81E+04 1.00E+1I 3.OOE+00 to 4.OOE+00 1.24E+10 1.54E-09 1.71E-13 2.18E-12 1.24E+10 4.OOE+00 to 6.OOE+00 6.85E+06 O.OOE+00 O.OOE+00 O.OOE+00 6.85E+06 6.OOE+00 to 8.OOE+00 7.88E+05 O.OOE+00 O.OOE+00 O.OOE+00 7.88E+05 8.00E+00 to 1.1OE+O1 9.07E+04 O.OOE+00 O.OOE+00 O.OOE+00 9.07E+04 Total 7.40E+15 3.68E+13 9.58E+12 1.31E+13 7.45E+15 Notes:

(1) Gamma Source is provided by Constellation Energy, Nuclear Generation Group in [2.4.9].

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 I PROJECT NO: 10955 PAGE: 19 of 67 Zone 3 Zone 3 Zone 3 Zone 3 Zone 3 Zone 3 Zone 1 Zone 1 Zone I Zone 1 Zone 3 Zone 3 Zone 3 IZone 3 Zone 3 Zone 3 Heat Zone Level No of FA kW/FA Total 1 4 0.8 3.2 2 8 1.0 8.0 3 12 1.0 12.0 4 8 0.8 6.4 Total Heat Load, kW 29.6 Figure 4-1 Heat Load Zone Configuration for the Maximum Heat Load Note: Four zones are employed to denote an increasing importance to dose rates, where Zone I represents the least important zone and Zone 4 represents the most important zone.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: UH32PHB.0 101 REVISION: 4 PROJECT NO: 10955 PAGE: 20 of 67 5.0 GENERAL DESIGN REQUIREMENTS The general requirements of the NUHOMS 32PHB canister are listed below. Specific component requirements are provided in subsequent sections.

The NUHOMS 32PHB canister shall meet the following requirements:

The "on-the-hook" weight limit shall not exceed the 125 ton crane limit. Draining of the cavity prior to removal from the pool may be used to meet the weight limits. During water removal a Helium or Nitrogen gas blanket is used to avoid fuel exposure to air.

The maximum per assembly heat load shall be 1.0 kW. The maximum heat load per canister shall be 29.6 kW. Zoning is used to accommodate total DSC heat loads up to 29.6 kW.

Air circulation in the TC/DSC annulus may be used to maintain cladding within the allowed temperature limits of ISG-1 I Rev. 3. The time limit to initiate air circulation following drainage of the TC/DSC annulus shall be greater than or equal to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months for the maximum heat load. The heat load zone configuration for the maximum heat load is shown in Figure 4-1. In addition to evaluation of the design basis heat load, the design shall also identify maximum heat loads with uniform heat load zoning configurations for the following conditions:

o Time limit to initiate air circulation is greater than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months o Time limit to initiate air circulation is greater than or equal to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

Average fuel assembly bumups up to 62 GWd/MTU to bound the data shown in Table 4-2.

" Maximum 5.0 wt.% initial U-235 fuel enrichment. Minimum 2.0 wt % U-235

Boron loading in the pool as function of U-235 enrichment (between 2450 and 2550 ppm)

The CCNPP-FC TC shall contain similar features, such as the spacer disc, airflow adaptor, and slotted lid, required for air circulation as in the OS197FC TC.

The 32PHB DSCs shall be stored only in the HSM-HB.

The CCNPP-FC TC shall be used to transfer the 32PHB DSC from the Auxiliary Building to the HSM-HB site.

The criteria and evaluations to be performed for the 32PHB DSC are similar to those used for the 32P documented in [2.4.3].

5.1 System Design Features and Considerations This section describes design features and considerations for the 32PHB Canister, the HSM-HB and the CCNPP-FC transfer cask.

5.1.1 Canister Features and Considerations The 32PHB canister shall meet the criteria below:

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.010OlREISION: 4_

PROJECT NO: 10955 PAGE: 21 of 67

Store up to 32 intact CE 14x14 Standard or VAP and/or AREVA 14x14 PWR spent fuel assemblies.

" For criticality control poison sheets shall be incorporated into the basket design. No structural credit shall be taken for the poison sheets.

" Nominal external diameter of 67.25" and a nominal shell thickness of 0.63".

Nominal canister length of 176.50", including the grapple ring on the outer bottom cover plate.

Nominal internal cavity length of 159".

Carbon (or stainless) steel end shield plugs with a lead core at both ends to reduce occupational dose levels.

Vent and siphon block shall be welded and integrated to the top shield plug assembly, similar to the HD license designs [2.4.13].

The inner diameter of the vent and siphon ports shall be approximately 1 inch.

" All carbon steel surfaces exposed to the fuel pool shall be electroless nickel coated to minimize the effects of hydrogen generation during fuel loading operations.

Proper interface shall be provided with the CCNPP-FC transfer cask and HSM-HB.

5.1.1.1 Weight Requirements Limit the maximum lifted load to less than 125 tons for the worst lift configuration with the heaviest fuel type. The maximum under-the-hook weight shall consider the following lift configurations:

Lift 1-from Decon Area to Fuel Pool Lift I consists of the transfer cask with an unloaded DSC. Both the DSC and the Cask/DSC annulus are filled with water. The cask is lifted from the cask decon area to the spent fuel pool. Its configuration is as follows:

Transfer Cask

- Top cover plate assembly removed

- Ram access cover plate installed

DSC

- Internal basket assembly installed

- Shield plug removed

- Top cover plate not installed

" Lifting yoke assembly with or without top shield plug attached and yoke extension slings (as applicable)

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: _4 PROJECT NO: 10955 ý PAGE: 22 of 67

Demineralized water in cask/DSC annulus (includes water above DSC)

Borated water in DSC cavity (partially displaced by basket assembly)

Lift 2-from Fuel Pool to Decon Area Lift 2 consists of the transfer cask and DSC loaded with spent fuel assemblies. Both the DSC and the Cask/DSC annulus are filled with water. The cask is lifted from the spent fuel pool to the cask decon area. Its configuration is as follows:

Transfer Cask

- Top cover plate assembly removed

- Ram access cover plate installed

" DSC

- Internal basket assembly installed

- Top shield plug installed

- Top cover plate not installed

Fuel assemblies loaded

Lifting yoke assembly and yoke extension slings (as applicable)

Demineralized water in cask/DSC annulus (includes water above DSC)

Borated water in DSC cavity (partially displaced by basket assembly and fuel assemblies)

Lift 3-from Decon Area to Transfer Trailer Lift 3 consists of the transfer cask and DSC loaded with spent fuel assemblies. Both the DSC and the Cask/DSC annulus are dry. The cask is lifted from the cask decon area to the transfer trailer. Its configuration is as follows:

Transfer Cask

- Top cover plate assembly installed

- Ram-access cover-plate installed

- Annulus is vented to atmosphere.

DSC

- Internal basket assembly installed

- Top shield plug installed

- Top cover plate installed

Fuel assemblies loaded

Lifting yoke assembly without yoke extension slings

Cask/DSC annulus and DSC cavity drained. Annulus is vented to atmosphere

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 23 of 67 5.1.2 HSM-HB Features and Considerations 5.1.2.1 HSM-HB Transfer Operation Features

Provide a round access opening compatible with the canister nominal outer diameter of 67.25 inches (32PHB).

The access opening horizontal centerline shall be nominally 8 feet 6 inches above ground to properly interface with the NUHOMS transfer equipment.

Provide a shielded door design (similar to HSM-H, Type B) that facilitates handling during canister transfer operations. Thickness and composition shall be established to meet the shielding requirements of Section 12.0.

" Interface with the modified on-site transfer cask (CCNPP-FC TC). Provide a means to restrain the transfer cask during canister insertion/withdrawal operations. This shall consist of embedded anchors capable of carrying a load of 55 kips (each side).

Provide a recessed cask 'docking surface to shield the end of the cask during canistir transfer.

Provide cask alignment targets on the exterior surface of the front wall on the cask horizontal and vertical centerlines.

Incorporate a hardened stainless steel surface into the support structure to facilitate canister sliding and minimize gouging during transfer.

5.1.2.2 HSM-HB Storage Operation Features

Provide support for a canister with a nominal outside diameter of 67.25 inches and nominal length of 176.50 inches. The HSM-HB design shall be based on a bounding dry loaded weight of 95 kips per canister.

" Provide seismic restraints to transfer horizontal loads during a seismic event to the concrete structure.

Provide structural resistance to protect the DSC from blast explosion.

5.1.2.3 HSM-HB Thermal Performance Features Provide inlet and outlet ventilation openings for passive airflow through the module interior.

Ambient air enters at the bottom and flows around the canister and exits at the top. The vent locations shall minimize radiation streaming and optimize airflow. The vent design shall minimize the possibility of becoming blocked or exposed by relative sliding of adjacent HSM-HBs under the effects of a dynamic event such as a seismic event, or tornado wind and missiles.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 24 of 67

" Protect vent openings with screens to prevent debris accumulation or animals from blocking or entering the HSM-HB.

" Provide the means to protect concrete surfaces closest to the canister from direct radiant heat to maintain general and local area temperatures in accordance with the requirements of Section 5.1.2.4.

Provide the means to monitor concrete temperature through the use of thermocouples embedded in the area where maximum concrete temperatures are expected during normal operation.

5.1.2.4 HSM-HB Material Criteria

All concrete used in the HSM-HB components shall be reinforced regardless of the functional role or need for structural strength or integrity (e.g. shielded door filling).

Reinforcing steel shall conform to ASTM A615 or A706, Grade 60 unless otherwise specified and approved in the design drawings. All concrete reinforcement shall have a minimum specified yield strength of 60 ksi.

Structural concrete used for the design shall have a 28 day specified compressive strength (f'c) of 5,000 psi.

" The concrete temperature limits criteria of Section A.4 Appendix A of ACI 349-97 [2.1.13] and those given in [2.4.16] shall apply.

If concrete temperatures resulting from thermal analyses of the HSM-HB result in temperatures that exceed the limits of Appendix A of [2.1.13], or those given in page 3-5 of [2.4.16], concrete testing will be required to demonstrate that the elevated temperatures do not reduce concrete strength below the values assumed in the HSM-HB structural analyses. Analyses are to be performed assuming a 10% reduction in concrete and rebar strengths at temperature. This reduction in strength will be used as the acceptance criteria for any testing performed. Typical procedures for concrete testing are provided in Appendix A of [2.4.19].

" The nominal density of reinforced concrete shall be assumed to be 150 pcf.

Fabrication of all miscellaneous steel framing, HSM-HB doors and embedments, heat shield, and screens shall conform to the applicable provisions of AISC Specification for Structural Steel Building [2.1.16] unless otherwise specified and approved in the design drawings. All welding shall be performed and welders qualified in accordance with the requirements of AWS D1.1

[2.1.17] or AWS D 1.6 [2.1.16], as appropriate.

No organic coatings on heat shields are permitted.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.001 REVISION: 4 PROJECT NO: 10955 PAGE: 25 of67 5.1.2.5 HSM-HB Geometry, Technical Specifications, and Weight Limits

" The HSM-HB configuration shall be based on optimizing the use of materials to achieve the stated dose limits in Section 12.0 with the flexibility to increase concrete section thickness to achieve more stringent criteria for site specific dose limitations. The design shall consider and qualify the enveloping geometry and weight.

The HSM-HB cavity height, cavity width, elevation and cross sectional areas of the HSM-HB air inlet/outlet vents, the total outside height, length, and width of the HSM-HB may not be deviate by more than 8% of their nominal design values as approved by the NRC on the drawings in [2.4.12], Appendix P.

The HSM-HB concrete shall be tested at elevated temperatures to verify that there are no significant signs of spalling or cracking and that the concrete compressive strength is greater than that assumed in the structural analysis. Tests shall be performed at or above the calculated concrete peak temperature and for a period no less than 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months (duration of HSM-HB blocked vent accident), if the calculated concrete temperature for accident conditions exceeds 350'F.

Typical procedures for concrete testing are provided in Appendix A of [2.4.19].

The precast HSM-HB components shall be designed to be transported by barge, truck or rail.

5.1.3 CCNPP-FC TC Features and Considerations As described in Section 3.0, the 32PHB DSC uses the modified CCNPP TC designated as CCNPP-FC TC for transfer operations. The CCNPP-FC TC has a top lid that allows cooling air, which is forced in at the bottom end of the cask to exit through cutouts around the perimeter of the TC top lid.

The CCNPP-FC transfer cask shall meet the criteria below:

To improve thermal performance, the TC is to allow forced cooling air to enter through the ram access cover plate opening. The ram access cover plate is replaced by an adapter cone that mates the fan hose to the ram access opening.

" A spacer with wedge shaped protrusions installed at the cask bottom plate is used to duct the airflow to the perimeter of the TC/DSC annulus.

Minimize pressure drop as the forced airflow is allowed to flow in the annulus between the DSC and the cask's inner shell.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955l PAGE: 26 of 67 6.0 ENVIRONMENTAL CONDITIONS This section provides the generic environmental conditions to be used in generating the design basis loads for the new components used in NUHOMS 32PHB system: the NUHOMS 32PHB DSC and the CCNPP-FC TC. The design basis loads for the transfer casks described in [2.4.141 are applicable to CCNPP-FC TC. The design basis loads for the HSM-HB module are the same as those described in the UFSAR [2.4.12], Appendix P for HSM-H. The loads considered are as follows:

Dead Weight

Wind and Tornado

Snow and Ice

" Seismic

Flood

Forest Fire

Thermal 6.1 Dead Load Dead load is the weight of the structure and attachments including permanently installed equipment.

For the CCNPP-FC TC analysis, the dead load shall be varied by + 5% if that produces the most adverse loading condition, regardless of the load combination factor applied.

The following deadweight loads shall be considered for the DSC:

1. The weight of the empty DSC (DWI), hanging vertically by the DSC lifting fixture without the Top Shield Plug and Top Cover Plate in place, shall be considered. The DSC stresses for this condition shall not exceed ASME code allowables with a load factor of 1.5.

2. The weight of the DSC (DW 2), including the Top Shield Plug, loaded with fuel and filled with borated water (hydrostatic head), resting in the Transfer Cask (TC) cavity with annuls water in vertical orientation, shall be considered.

3. The weight of the dried DSC (DW 3 ), including the top shield plug and top cover plate, loaded with fuel resting horizontally in the Transfer Cask cavity or the HSM-HB, shall be considered.

6.2 Wind, Tornado, and Snow There are no credible wind, tornado, or snow loads applied to the canister as the HSM-HB and CCNPP-FC TC provide the environmental protection.

The design basis loads related to wind, tornado, and snow are described in [2.4.14], Section 4.3 are applicable to CCNPP-FC TC. The design loads for HSM-HB are the same as those described in

[2.4.12], Appendix P, Section P.2.2.1 for HSM-H.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PRB.0101 TREVISION: 4 PROJECT NO: 10955I PAGE: 27 of 67 6.3 Seismic The seismic loads induced into DSC through the HSM-HB or TC shall be developed in accordance with NRC Regulatory Guides 1.6 and 1.61 ([2.3.3] and [2.3.4] with ground accelerations of 0.15g horizontal and 0.1Og vertical, with 3% critical damping. In lieu of a site specific seismic analysis of the HSM-HB/DSC support structure and loaded DSC, the DSC assembly may be analyzed for the above seismic accelerations applied as equivalent static loads in accordance with Section 8.2.3.2, Paragraph A.ii, "DSC Seismic Stress Analysis", of the NUTECH Topical Report NUH-002 [2.4.1],

If it is decided to perform a dynamic response spectrum analysis for seismic loads, a procedure consistent with NRC Regulatory Guide 1.92 [2.3.6] shall be used for combining the response values for individual modes.

The seismic loads described in [2.4.14] are applicable to CCNPP-FC TC. The seismic loads for HSM-HB are the same as those described in [2.4.12] for HSM-H and remain unchanged.

6.4 Flood Flood loading is excluded by the ISFSI USAR and need not be considered in the design of the 32PHB DSC.

6.5 Fire Accident A postulated fire accident shall be considered based on ISFSI USAR [2.4.5], Section 3.3.6. A maximum fuel spill of 100 gallons of diesel fuel, which is the maximum capacity of both fuel tanks within the tow vehicles, shall be considered for this evaluation.

6.6 Forest Fire A postulated forest fire shall be considered based on ISFSI USAR [2.4.5]. The forest fire shall be assumed to occur at a distance of 130 feet from the nearest HSM-HB. The flame front is 200 feet long by 100 feet in height burning at an effective flame temperature of 1832°F for a period of I hour. The flame emissivity is 0.9. An average initial concrete temperature may be considered for this analysis based on off-normal ambient conditions described in Section 6.7.2.

6.7 Thermal Environmental Conditions 6.7.1 Fuel Handling Building Conditions Based on CCNPP specification [2.4.2], Section 3.3, the maximum ambient temperature for all operations is 104'F. To account for the heat up of the Auxiliary Building during the loading operation, the loaded canister and CCNPP-FC TC will be analyzed for a normal (steady state) average ambient temperature of 100'F for operations which take place inside the Auxiliary Building, including vacuum drying, blowdown, and canister cavity water heat up. The maximum pool water temperature is 140'F. Maximum relative humidity is 100%.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 T.REVISION: 4 PROJECT NO: 10955l PAGE: 28 of 67 6.7.2 ISFSI Site Conditions Thermal loading for the CCNPP-FC TC and HSM-HB shall consider the full range of plausible natural weather temperatures and fluctuations and the heat dissipation from the stored canisters. As a minimum, the following conditions shall be used in the appropriate thermal analyses.

Off-normal ambient temperature range of -87F without insolation to 1047F with full insolation. A solar heat flux of 127 Btu/hr-ft2 on the outer surface of the HSM-HB shall be assumed to occur concurrently with the 1047F ambient.

Normal ambient temperature range of-8°F without insolation to 1047F with full insolation.

A solar heat flux of 82 Btu/hr-ft 2 on the outer surface of the HSM-HB shall be assumed to occur concurrently with the 1047F ambient. The design lifetime average ambient temperature is 707F based on ISFSI USAR [2.4.5], Section 8.1.3.

Relative humidity of 100%.

" Blockage of the inlet or outlet cooling vents at the extreme ambient temperatures of-8°F and 1047F. A solar heat flux of 127 Btu/hr-ft2 on the outer surface of the HSM-HB shall be assumed to occur concurrently with the 104'F ambient. The maximum time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months shall be considered for the blocked vent transient with the maximum heat load for the design basis fuel.

6.8 Other Natural Phenomena Other natural phenomena, such as lightning, tsunamis, and hurricanes are described in ISFSI USAR

[2.4.5]. The effects of these site specific phenomena are generally bounded by other events and are excluded from the scope of this criteria document.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0 01 TREVISION: 4 PROJECT NO: 10955 PAGE: 29 of 67 7.0 OPERATIONAL CONDITIONS 7.1 HSM-HB The design criteria for the HSM-HB are the same as those documented in the Design Criteria Specification for the NUHOMS 32PTH 1 System, Reference [2.4.14] for HSM-H except for the items defined in Section 7.1.1 below.

7.1.1 Normal and Off-Normal Operational Handling Loads Normal operation handling loads on the HSM-HB are the result of canister transfer operations.

Normal operation assumes the canister is sliding over the support structure due to a hydraulic ram force of up to 23,750 lbs applied at the grapple ring [2.4.4].

The design basis off-normal operating load is due to a hydraulic ram force of 80,000 lbs applied at the grapple ring.

The axial load should be transferred to the HSM-HB support structure. In addition, half the loaded weight of the canister should be applied as a concentrated load midspan of the HSM-HB support structure.

The design loads for HSM-HB structure are listed in Table 8-11. The design loads are bounding for the handling forces from the hydraulic ram.

7.2 32PHB Canister The canister loads are developed for normal, off-normal, and accident conditions. The following provides a general discussion of the operational loads. Detailed criteria are provided in Section 8.0.

7.2.1 Normal Operational Loads Normal operational loads are defined as the dead loads, pressure and temperature conditions that result from normal storage and transportation conditions. Normal operational handling loads result from the transfer of a canister from the fuel building to the HSM-HB using the CCNPP-FC on-site transfer cask. The loads considered shall include those transferred from cask to canister and the insertion/withdrawal loads during transfer in/out of an HSM-HB. These loads are described in detail in Section 8.0.

7.2.2 Off-Normal Operational Loads Off-normal operation handling loads are the result of a canister getting stuck or jammed during transfer into or out of the HSM-HB. These loads are described in detail in Section 8.0.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 30 of 67 7.2.3 Accident Operational Loads These loads are the result of a loaded 32PHB canister drop event associated with the CCNPP-FC transfer cask events. For transfer operations the canister is protected from the environment and operational loads by the TC and the canister provides the confinement boundary for the fuel. The cask/canister drop events are arbitrarily defined with no mechanistic basis for the loads.

7.3 CCNPP-FC Transfer Cask To limit the off-gas pressure within the neutron shield panel and limit the hydrogen loss of the NS-3 neutron shield to less than 10%, a maximum bulk temperature of 280'F is considered for the NS-3 neutron shield. The set point of the TC neutron shield safety relief valve is 95 psig according to

[2.4.5], Section 8.1.3.3.

The structural design criteria for the CCNPP-FC TC are described in Section 9.0.

7.4 Load Combinations Individual load conditions are listed in Table 7-1. Summaries of the load combinations are shown in Table 7-2 and Table 7-3 for DSC shell and basket assemblies, respectively. These tables include the applicable ASME service level for each combination. Analyses for the on-site "accident" conditions may use either elastic or elastic-plastic analyses with the appropriate allowables from Table 8-7 and/or Table 8-8.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 31 of 67 Table 7-1 DSC Structural Loading Conditions Item No. Applicable Individual Loads Condition Load Value I Dead Weight ......

2 Hot ambient of 104 0 F ---

3 Temperatures Cold ambient -8' F ---

4 Transfer Accident ---

5 HSM-HB Accident ---

6 Normal 15 psig 7 Internal Pressure Off-Normal 20 psig 8 Accident 100 psig 9 Hydrostatic Pressure (External) ......

10 Normal I1 Handling Loads Off-Normal See Table 8-11 12 Accident 13 Dead load +/- I g vertical 14 Normal Dead load +/- Ig axial 15 Dead load +/- ig longitudinal 16 Transfer Loads Dead load +/- 1/2/g all directions 17 Top End Drop, 75g 18 Accident Bottom End Drop, 75g 19 Side Drop, 75g 0.36g Axial + 0.36g Axial +

20 0.41g Transverse + 0.41g Transverse +

Seismic 0.25g Vertical + 1.25g Tran 1.Og Down (DW) 1.25g Vertical Stability 21 0.41 g Transverse +

0.25g Vertical 22 Test Pressure Normal 16.5 psig 23 Blow-Down Pressure Normal 20 psig

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 REVISION: 4 PROJECT NO: 10955 PAGE: 32 of 67 Table 7-2 Summary of 32PHB DSC Shell Load Combinations Normal Operating Conditions Off-Normal Conditions Emergency Conditions / Accident Load Combinations Case Conditions 1 2 3 4 5 6 7 1 2 3 4 5 6 1 2 3 4 5 6 7 Vertical, DSC Empty X Dead Weight Vertical, DSC w/ and w/o Water X Horizontal, DSC w/Fuel X X X X X X X X X X X X X X X X X X Inside HSM: 701F (ambient) X X Inside Cask: 70°F (ambient) X X X X X X X Thermal Inside HSM: 104°F (ambient) X X X Inside Cask: 104°F (ambient) X X X Inside Cask: Accident X X Inside HSM: Accident (vent block) X Normal Pressure X X X X X Internal Pressure Off-Normal / Blowdown X X X X X X X Accident X X External Pressure Hydrostatic X Normal X X -

Hand ling Lo ads ----------

Off-Normal X X X Normal X X X X Transfer Loads Accident (Drop) X X Seismic X ASME Code Service Level A A A A A A A B B B B B B C C D D D D D

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB.0101 I REVISION: 4 I PROJECT NO: 10955 1 PAGE: 33 of 67 Table 7-3 Summary of 32PHB Basket Load Combinations Normal Operating Conditions Off-Normal Conditions Emergency Conditions /

Load Combinations Case _ Accident Conditions 1 2 (2) 3 (3) 1 2 1 2 3 Vertical X Dead Weight Horizontal X X X X X X X Inside HSM: 70*F (ambient) X Inside Cask: 70°F (ambient) X Thermal Inside HSM: 104*F (ambient) X X Inside Cask: 104°F (ambient) X X Inside HSM: Accident (vent block) X Normal X X T ra ns fer Lo a d s Acc d en *...

Accident X_ _

Seismic X ASME Code Service Level A A A B B C D D Notes:

( Side Drop orientations of 00, 450, 600, and 1800, and End Drop should be considered for accident transfer load analysis.

(2) This load case is bounded by Off-Normal Condition load case 1.

('3) This load case is bounded by Off-Normal Condition load case 2.

A TRANSNUCLEAR AN AREVA COMPANY

ýDOCUMENTNO: NUH32PHB-0101 __[REVISION: 4 PROJECT NO: 109551 PAGE: 34 of 67 8.0 DSC STRUCTURAL DESIGN REQUIREMENTS 8.1 NUHOMS 32PHB DSC Structural Design Criteria The 32PHB DSC is designed to meet the criteria of ASME Code Subsection NB. Service Level A and B allowables are used to for all normal operating and off-normal loadings. Service Level C and D allowables are used for load combinations that include postulated accident loadings.

Material Properties used in the 32PHB analysis are summarized in Table 8-1 through Table 8-6.

The material properties for lead and air are listed in Table 9-7 through Table 9-9.

8.1.1 NUHOMS 32PHB DSC Shell Stress Limits The stress limits for the DSC shell are taken from the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, Article NB-3200 for Level A through D Service Limits [2.1.2]. In accordance with NB-3225, Appendix F is used for accident condition loads (Level D).

The stress due to each load shall be identified as to the type of stress induced, e.g. membrane, bending, etc., and the classification of stress, e.g. primary, secondary, etc.

Stress limits for Level A through D service loading conditions are summarized in Table 8-7. Local yielding is permitted at the point of contact where Level D load is applied. If elastic stress limits cannot be met, the plastic system analysis approach and acceptance criteria of Appendix F of Section III shall be used.

The allowable stress intensity value, Sm, as defined by the Code, is to be based on the calculated (or a bounding) temperature for each service load condition.

The canister closure welds shall be designed in accordance with the guidance of ISG-15 [2.3.23].

Structural stability of the 32PHB DSC shell assembly is evaluated for those load conditions in which the DSC is under external pressure and/or axial compression. The stability criteria for Level A load conditions are from ASME Section 111, NB-3133.3 and NB-3133.6 [2.1.2]. For the accident drop the shell assembly stability load may be evaluated based on non-linear, large displacement analysis model using ANSYS.

8.1.2 NUHOMS 32PHB Canister Basket Stress Limits The basket fuel compartment tube wall thickness is established to meet heat transfer, nuclear criticality, and structural requirements. The basket structure must provide sufficient rigidity to maintain a subcritical configuration under the applied loads.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NLJH32PHB-0101 REVISION: 4 PROJECT NO: ]10955I PAGE: 35 of 67 The stress analyses of the basket do not take credit for the poison plates except for through thickness compression. However, the poison plate strength may be considered when determining secondary stresses in the stainless steel.

The basis for the stainless steel fuel compartment section stress allowables is the ASME Code,Section III, Subsection NG [2.1.2]. Stress limits for Level A through D service loading conditions are summarized in Table 8-8.

Alternatively, and in accordance with NG-3222 and Note 9 to Figure NG-3221-1, the Limit Analysis provisions of NG-3228 may be used for Level A Service Limits.

The basket shall be evaluated under Level D Service loadings in accordance with the Level D Service limits for components in Appendix F of Section III of the Code. The hypothetical impact accidents are evaluated as short duration Level D conditions. For elastic quasistatic analysis, the primary membrane stress is limited to the smaller of 2 .4 Sm or 0.7S, and membrane plus bending stress intensities are limited to the lesser of 3.6 Sm or S". The maximum primary shear stress is limited to 0.42 S,. When evaluating the results from the non-linear elastic-plastic analysis for the accident conditions, the general primary membrane stress intensity, Pm, shall not exceed the greater of 0.7Su or Sy + 1/3 (Su - Sy) and the maximum stress intensity at any location (PI or P, + Pb) shall not exceed 0.9 S,.

Finite element non-linear buckling analysis or hand calculations should be used in calculating the critical loads for buckling of the shell and basket. Reasonable safety factors for the allowable buckling loads should be provided to take into account material and geometrical imperfections.

The solid aluminum transition rails perform their function by remaining in place. The loads on these rails are primarily bearing. Therefore, for deadweight and handling conditions, stress in the solid aluminum bodies will be compared to the allowable bearing stress, Sy, from NG-3227.1(a). For accident condition loads (i.e. postulated drops), the rail bodies support the fuel compartment tube structure such that stresses and displacements in the compartment tube structure are acceptable.

Since the solid aluminum rails are entrapped between the fuel compartment tube structure and the DSC shell, no additional checks of the aluminum are required for accident loading. Qualification of the fuel compartment tube structure demonstrates that the rails perform their intended function.

The fuel to be stored in the 32PHB DSC is described in Section 4.0. Under side loads (e.g.,

horizontal dead weight, side drops, etc.), the fuel is assumed to have no structural capacity. That is, the fuel load is applied to the basket fuel compartments as a pressure load which varies with axial location of the fuel and components along the 32PHB DSC cavity.

The load imposed by the fuel on the basket fuel compartments is a function of the linear weight distribution (lb/in) of the fuel assemblies. That is, the load in the active fuel region may be larger than the load at the ends of the fuel assembly. This linear weight distribution shall be determined using the weight and geometry data given in Section 4.0. Alternatively, a conservative bounding weight per linear inch may be used.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-01 01 REVISION: 4 PROJECT NO: 10955 1PAGE: 36 of 67 The criteria for the fusion welds in the basket shall be by test. These are resistance (fusion) welds, and are not addressed explicitly in NG-3352-1. This is a Code exception to NG-3352-1. Fusion welds between the stainless steel support bars and the stainless steel fuel compartments shall be qualified by testing. The required minimum tested capacity of the weld connection shall be based on the margin of safety (test to design) of 1.43 (see Appendix F, Section 1342(c) of ASME code

[2.1.4]), corrected for the temperature difference between testing and basket operating conditions and the maximum weld load at any weld location in the basket.

The stress and load limits for the basket are summarized in Table 8-8.

8.2 Fuel Assembly Evaluations The fuel assemblies are evaluated to demonstrate fuel cladding integrity under normal, off-normal, accident, loading, and unloading conditions. The evaluations may address individual assemblies or may be based on a single bounding assembly. The evaluations may use hand calculations or computer analysis models. Equivalent static or dynamic analysis may be used. In accordance with ISG-12 the weight of the pellets is included in the analysis. However, no credit for the pellets is to be taken in the development of the section properties of the cladding (cross sectional area, moment of inertia). Cladding material properties are to be consistent with high burn up fuel and should include a thickness reduction due to oxidation.

The analysis should use irradiated material properties and strain rate effects as per References

[2.4.25] and [2.4.20]. The acceptance criterion is per ISG-1 2, i.e., cladding integrity is assured if the cladding stress remains below yield strength. Alternatively the methodology and criteria from Reference [2.4.23] may be applied. Material properties may be based on Reference [2.4.24].

8.3 Weld stresses The 32PHB DSC closure welds are in accordance with the guidance of ISG-I 5 [2.3.23]. These include the joints between the top cover plate and top shield plug assembly to the shell and the vent and siphon block welds in the top shield plug asembly. These welds are partial penetration welds subject to PT examination. Other DSC shell assembly welds are in accordance with NB-4243 and NB-5230.

8.4 Canister Loads Descriptions A summary of the load and load combinations used for the 32PHB design was listed in Table 7-2.

This section provides a summary description of the individual loads requiring evaluation for 32PHB DSC design.

8.4.1 Deadweight Deadweight loads include the self-weight of the 32PHB DSC (including basket assembly), and stored fuel.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 R V ISIO0N:

RE _4 PROJECT NO: 10955[ PAGE: 37 of 67 8.4.2 Internal Pressure Pressure loads for the 32PHB DSC are based on conditions defined by NUREG- 1536, Section 4.0.V.5.c and NCA-2142.1 (a).

For storage considerations, it shall be assumed that 1% of the fuel rods are failed for normal conditions, up to 10% of the fuel rods are failed for off normal conditions, and 100% of the fuel rods will have failed following a design basis accident. Since no drop is postulated for the HSM-HB blocked vent accident condition, it should be assumed that only 10% of the fuel rods are failed for this accident condition.

The total amount of fission and filled gases that can be released from each fuel rod is 97.9186 in3 at STP (68'F and 1 atm) per fuel rod for 62 GWD/MTU burnup.

Required pressures for structural analysis for Normal, Off-Normal, and Accident conditions are listed in Table 8-9. The helium backfill pressure of the DSC is defined as 2.5+/-1.0 psig.

8.4.3 Thermal For the 32PHB DSC analyses, temperature profiles and maximum component temperatures are based on thermal analyses, which consider the environmental conditions listed in Table 8-10 with a maximum heat load of 29.6 kW per DSC and heat load zoning configuration shown in Figure 4-1.

The component temperatures shall be used in determining the allowable stresses for each condition.

Maximum thermal gradients are to be considered for determining thermal stresses.

8.4.4 Seismic As described in Section 6.3, the 32PHB seismic criteria consists of the Regulatory Guide 1.60 Response Spectra, anchored to 0.1 5g and 0.1 Og horizontal and vertical peak ground accelerations, respectively. The 32PHB DSC shall be analyzed for seismic accelerations considering the dynamic response of the HSM-HB, as applicable. In addition, the 32PHB DSC shall be analyzed considering the possibility of loading only a single rail. The seismic loading is considered by simultaneously applying the seismic accelerations as equivalent static loads in each orthogonal direction. The results for each direction may be combined by SRSS. For a seismic event occurring during transfer of a 32PHB DSC, the transfer loads specified in Section 8.4.5.1 provide assurance for the integrity of the 32PHB DSC components.

8.4.5 Handling There are two categories of "handling" loads: (1) inertial loads associated with moving the 32PHB DSC and (2) the loads associated with inserting the 32PHB DSC into (and retrieving the 32PHB DSC from) the HSM-HB.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 FREVISION: 4 PROJECT NO: 10955 PAGE: 38 of 67 8.4.5.1 Transfer Handling Transfer handling loads are inertial loads on the loaded 32PHB DSC resulting from on-site handling and transportation between the fuel handling/loading area and the HSM-HB. The four independent load cases to be considered in the design are:

+/- 1.Og Axial

+/- 1.Og Transverse

+/- I.Og Vertical

+/- 0.5g Axial +/- 0.5g Transverse +/- 0.5g Vertical (cask horizontal) 8.4.5.2 HSM-HB Insertion/Retrieval To load the 32PHB DSC into the HSM-HB, the 32PHB DSC is pushed out of the transfer cask using a hydraulic ram. The load is applied to the center of the 32PHB DSC outer bottom cover at the center of the grapple ring assembly.

To unload the HSM-HB, the 32PHB DSC is pulled using grapples which fit into the grapple ring on the outer bottom cover. The allowable hydraulic ram forces and the design loads considered for Insertion/Retrieval loads are listed in Table 8-11. It should be noted that the design loads are bounding for the hydraulic ram forces.

8.4.6 Drop Loads Postulated on-site transfer drop loads apply to all parts of the 32PHB DSC. The following loads are to be considered for on-site transfer drops.

Equivalent static deceleration:

7 5 g vertical end drop 75 g horizontal side drop 2 5 g comer drop with slap down (corresponds to an 80" drop height)

Structural damping during drop: 10%

The above loads are taken from USAR [2.4.5] for NUHOMS 24P and 32P designs. No evaluation is required for comer drop since the stresses are bounded by the vertical drop stresses as shown in USAR [2.4.5].

8.4.6.1 Tornado Wind and Missile Loads The 32PHB DSC is protected from tornado wind and missile loads by the transfer cask or the HSM-IHB. Therefore, no evaluation is required for these loads.

8.4.7 Flood Loads Flood loading is excluded by CCNPP Specification 2.4.4 and need not be considered in the design of the DSC.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 TREVION 4 PROJECT NO: 10955 PAGE: 39 of 67 Table 8-1 Material Properties - SA-240/SA-479 Type 304 (DSC)

Nominal Composition: 18Cr-8Ni Temp E Sm SY Su (XINST OtAVG P k a (TF) (103 ksi) (ksi) 6 3 (ksi) (ksi) (I0 OF-1) (10-6 OF-) (lb/in ) (Btu/hr-ft-°F) (ft2/hr)

-20 20.0 30.0 75.0 70 28.3 8.5 8.5 8.6 0.151 100 20.0 30.0 75.0 8.7 8.6 8.7 0.152 150 26.7 9.0 8.8 200 27.6 20.0 25.0 71.0 9.3 8.9 9.3 0.156 250 23.6 9.6 9.1 300 27.0 20.0 22.4 66.2 9.8 9.2 9.8 0.160 350 10.0 9.3 400 26.5 18.7 20.7 64.0 10.2 9.5 10.4 0.165 450 10.3 9.6 500 25.8 17.5 19.4 63.4 10.4 9.7 0.290 10.9 0.170 550 10.6 9.8 600 25.3 16.4 18.4 63.4 10.7 9.8 11.3 0.174 650 16.2 18.0 63.4 10.8 9.9 700 24.8 16.0 17.6 63.4 10.9 10.0 11.8 0.179 750 15.6 17.2 63.3 11.0 10.0 800 24.1 15.2 16.9 62.8 11.1 10.1 12.2 0.184 850 16.5 62.0 11.1 10.1 900 23.5 16.2 60.8 11.2 10.2 12.7 0.189 950 15.9 59.3 11.3 10.3 1000 22.8 15.5 57.4 11.4 10.3 13.2 0.194 Table Table Table Table U SA-240 TM-I 2A Y-1 304 pg 606.1 pg 328 pg 520 pg 453.3 group G line 12 line 9 Table TE-1I Perry Calculated based on Table Table Table U pg 583 Table Table TCD Table 2A Y- T group 3 23-5 pg 594 SA-479 TM-I pg 328 pg 520 pg 453.3 group J 304 pg 606.1 line 26 line 25 line 25 group G ASME Section 11, Part D - Properties, 1998 with 1999 Addenda Perry Chemical Engineers' Handbook, 5 h Edition

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 _1REVISION: 4_

PROJECT NO: 10955l PAGE: 40 of 67 Table 8-2 Material Properties - Aluminum 6061 (DSC)

SB-209 A96061 T651 (.25" - 4.0") or T6 (.051" - .249")

Temp E Sm SY (XINST (XAVG P k a (OF) (10 ksi) (ksi) (ksi) (10-6 OF-) (10-6 TF-) (lb/in 3 ) (Btu/hr-fl-0 F) (ft2/hr)

-20' 14.0 35.0 70 10.0 12.1 12.1 96.1 2.66 100 14.0 35.0 12.5 12.4 96.9 2.66 150 14.0 34.6 12.9 12.7 98.0 2.65 200 9.6 14.0 33.7 13.3 13.0 99.0 2.65 250 13.4 32.4 13.6 13.1 0.098 99.8 2.64 300 9.2 11.3 27.4 13.9 13.3 100.6 2.63 350 20.0 14.2 13.4 101.3 2.62 400 8.7 13.3 14.6 13.6 101.9 2.62 450 1 14.8 13.8 500 8.1 14.9 13.9 1 550 15.2 14.1 600 15.3 14.2' Table Table Table Y-I Calculated based on TM-2 2A pg 552 Table TE-2 Table NF-2 Table TCD pg 607 pg 366 lines pg 585 pg 611,612 pg604,groupA96061 A96061 lines 3/4 23/24 pg 604, group A96061 ASME Section 11, Part D - Properties, 1998 with 1999 Addenda Table 8-3 Analysis Properties for Aluminum Transition Rails (DSC) [2.4.26]

6061-0 Aluminum (Annealed)

Temperature SO, 606 1-0 Sy, 6061 -0 E (OF) (ksi) (ksi) (10' ksi) 75 18.0 8.0 9.9 212 18.0 8.0 9.5 300 15.0 8.0 9.1 350 12.0 8.0 8.9 400 10.0 7.5 8.6 450 8.5 6.0 8.3 500 7.0 5.5 7.9 600 5.0 4.2 6.8 700 3.6 3.0 5.5 800 2.8 2.2 ---

900 2.2 1.6 1 1000 1.6 1.2 1 ---

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4_

PROJECT NO: 10955[ PAGE: 41 of 67 Table 8-4 Material Properties - Aluminum 1100 (DSC)

Aluminum Alloy A91 100 Temp E UINST LAVG P k (OF) (103 ksi) (106 oF-I) (10.6 OF-') (lb/in 3) (Btu/hr-ft-.F) (ft2/hr)

-20 70 10.0 12.1 12.1 133.1 3.67 100 12.5 12.4 131.8 3.61 150 12.9 12.7 130.0 3.50 200 9.6 13.3 13.0 128.5 3.42 250 13.6 13.1 127.3 3.35 300 9.2 13.9 13.3 0.098 126.2 3.28 350 14.2 13.4 125.3 3.23 400 8.7 14.6 13.6 124.5 3.17 450 14.8 13.8 500 8.1 14.9 13.9 550 15.2 14.1 600 15.3 14.2 Table TM-2 pg 607pg 607Table p58TE-2 Table NF-2TalTC Calculated TCD g6112Table based on A91100 pg585 pg611,612 pg 604, group A91100 ASME Section 11, Part D - Properties, 1998 with 1999 Addenda

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 I PROJECT NO: 10955 PAGE: 42 of 67 Table 8-5 Material Properties - Helium (DSC)

Temperature Thermal conductivity Temperature Thermal conductivity (K) (W/m-K) (7F) (Btu/hr-in-0 F) 300 0.1499 80 0.0072 400 0.1795 260 0.0086 500 0.2115 440 0.0102 600 0.2466 620 0.0119 800 0.3073 980 0.0148 1000 0.3622 1340 0.0174 1050 0.3757 1430 0.0181 The above data are calculated based on the following polynomial function from: Rohsenow, W.M., "Handbook of Heat Transfer", 3rd Edition, McGraw-Hill Handbooks.

k = _CT for conductivity in (W/m-K) and T in (K)

For 300 < T < 500 K for 500< T < 1050 K CO -7.761491E-03 CO -9.0656E-02 CI 8.66192033E-04 CI 9.37593087E-04 C2 -1.5559338E-06 C2 -9.13347535E-07 C3 1.40150565E-09 C3 5.55037072E-10 C4 0.OE+00 C4 -1.26457196E-13 Table 8-6 Material Properties - Nitrogen Temperature Thermal conductivity Temperature Thermal conductivity (K) (W/m-K) (OF) (Btu/hr-in-°F) 366 0.0304 200 1.466E-03 422 0.0304 300 1.636E-03 478 0.0373 400 1.797E-03 533 0.0405 500 1.950E-03 588 0.0435 600 2.096E-03 644 0.0464 700 2.236E-03 700 0.0493 800 2.372E-03 755 0.0520 900 2.503E-03 811 0.0546 1000 2.630E-03 866 0.0572 1100 2.753E-03 The above data are calculated based on the following polynomial function from: Rohsenow, W.M., "Handbook of Heat Transfer", 3rd Edition, McGraw-Hill Handbooks.

k =Z C, T' for conductivity in (W/m-K) and T in (K)

For 250 < T < 1050 K CO -1.52318E-03 C3 1.15568E-10 C1 1.1 8880E-04 C4 -6.36537E-14 C2 -1.20928E-07

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB-0101 I REVISION: 4 PROJECT NO: 10955 1 PAGE: 43 of 67 Table 8-7 Summary of Stress Criteria for Subsection NB Pressure Boundary Components (e.g., Shells and Cover Plates)

(continued)

Service Level Stress Category References Notes Pm :- 1.0S, PL < 1.5Sm Design Pm (or PL)+ Pb <1. Sm 5 NB-3221.1, NB-3221.2, NB-

- 3221.3, NB-3227.1 and

[NB-3221] Fp < 1.5Sy NB-3227.4 a 1 + 472 + U 3 <- 4 S, External Pressure: NB-3133 Pm - 1.0Sm PL < 1.5Sm Pm (or PL)+Pb < 1.5Sm Level A ( NB-3222, NB-3227.1, & Notes

[NB-32221 Pm (or PL)+Pb +Q:5 3.0S NB-3227.4 I&2 Fp :51.5Sy 4

a1 + ++2U-3 Sm External Pressure: NB-3133 Pm !5 1.0Sm PL 1.5S, Level B PL)+Pb 5 .5Sm NB-3223, NB-3227. 1, & Notes

[NB-3223] Pm (or PL)+Pb +Q 5 3.0Sm NB-3227.4 I&3 Fp <51.5Sy

+ C2 + U 3 5 4Sm

+l External Pressure: NB-3133 Pm <- max(1.2Sm, 1.0Sy)

PL <-max(l .8S,, 1.5Sy)

Level C Pm (or PL)+Pb *max(l.8Sm, 1.5SY) NB-3224, NB-3227.1 &

[NB-3224] Pm (or PL)+ Pb + Q 5 note 4 NB-3227.4 Fp <1.5Sy a1 + U 2 + U3 < 4 Sm External Pressure: 1.20*NB-3133

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 I REVISION: 4 ýI PROJECT NO: 10955 PAGE: 44 of 67 Table 8-7 Summary of Stress Criteria for Subsection NB Pressure Boundary Components (concluded)

Service Level Stress Category References Notes Carbon Steel Components (e.g., shield plugs)

Pm_ 0.7S, Level D Pm (orPL )+ Pb, 1. OSu Elastic Analysis N13-3225, F-1331.16 Ne

[NB-3225. App. F] Pm + +Q < note 4 & F-1331.5(b)

Fp <_note 5 External Pressure = 1.5

NB-3133 Pm < 0.7Su Level D Pm (orPL )+Pb < 0.9S" NB-3225, F-1341.2 Plastic Analysis Pm+ Pb + Q! <note 4 & F - 1331.5(b) Note 6

[NB-3225,. App. F] F <note 5 External Pressure = 1.5

NB-3133 Austenitic Steel Components (e.g., Shell)

Pm < rnin(2.4Sm, 0.7S,,)

Level D Pm (orPL)+Pb <min(3.6Sm,1 .0Su) NB-3225, F-1331.1 Elastic Analysis Pm + P + Q < note 4 & F - 1331.5(b) Note 7

[NB-3225, App. F] Fp* note 5 External Pressure = 1.5

NB-3133 Pm <max(0.7Su,Sy +(Su- Sy)/3)

Level D Pm (or PL )+Pb <0.9Su NB-3225, F-1341.2 Note 7 Plastic Analysis Pm+ Pb + Q:< note 4 & F-1331.5(b)

[NB-3225, App. F] m : note 5 &

Fp<note External Pressure = 1.5

NB-3133 Notes:

") This limit may be exceeded provided the criteria of NB-3228.5 are satisfied.

(2) There are no specific limits on primary stresses for Level A events. However, the stresses due to primary loads during normal service must be computed and combined with the effects of other loadings in satisfying other limits. See NB-3222.1.

(3) The 10% increase in allowables from NB-3223(a) may be applicable for load combinations for which the pressure exceeds the design pressure.

(4) Evaluation of secondary stresses not required for Level C and D events.

(') Evaluation of bearing stresses not required for Level D events.

(6) Criteria listed are for carbon steel components (e.g., shield plugs).

(') Criteria listed are for austenitic parts including shells, cover plates, and the grapple assembly.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 T REVI1SION: _4 I

PROJECT NO: 10955I PAGE: 45 of 67 Table 8-8 Summary of Stress Criteria for Subsection NG Components (Austenitic)

(e.g., Fuel Compartment, Transition Rails)

Service Level Stress Category (5) References Notes Design Pm <-1.0Sm NG - 3221.1

[NG-3221] P,+ Pb < 1.5S, NG -3221.2 Pm <-1.0Sm NG-3222.1,NG-3221.1 Level A Pm+ Pb < 1.5Sm NG- 3222.1, NG- 3221.2 Note 6

[NG-3222] Pm+ Pb + Q <- 3."OS (note 4) NG- 3222.2 LevelB Pm <-1.0Sm NG- 3223(a), NG - 3222.1, NG - 3221.1

[NG.3223lB Pm +Pb 5 1.5Sm NG-3223(a)NG-3222.1,NG-3221.2 Note 1 Pm + Pb +QQ- 3.0Sm. (note 4) NG -3223(a),NG-3222.2 P.

1.5S. NG - 3224.1(a)(1)

Level C Notes Elastic Analysis P. + Pb < 2.25S. NG - 3224.1(a)(2)

[NG-3224] Pm + Pb + Q: <note 2 Figure NG - 3224 - 1 2 &3 Level D P5 *min(2.4S,0.7S,) NG - 3225, F - 1440, F - 1331.1(a)

Elastic Analysis P+ Pb !5min(3.6Sm, 1.os) NG - 3225, F - 1440, F - 1331.1(c)

[NG-3225, App. F] Pm + Pb + Q < note 2 P_ <max10.7S., Sy + 1/3(S. -Sy )

Level D NG - 3225, F - 1440, F - 1341.1(a)

Plastic AnalysisF] + Pb *09S. NG - 3225, F - 1440, F - 1341.2(c)

[NG-3225, App. F] Pm + Pb + Qnote 2 Notes:

(') There are no pressure loads on the basket, therefore the 10% increase permitted by NG-3223(a) for pressures exceeding the design pressure are not included.

(2) Evaluation of secondary stresses not required for Level C and D events.

(3) Criteria listed are for elastic analyses, other analysis methods permitted by NG-3224.1 are acceptable if performed in accordance with the appropriate paragraph of NG-3224.1.

(4) This limit may be exceeded provided the requirements of NG-3228.3 are satisfied, see NG-3222.2 and NG-3228.3.

O As appropriate, the special stress limits of NG-3227 should be applied.

(6) In accordance with NG-3222 and Note 9 of Figure NG-3221-1, the Limit Analysis provisions of NG-3228 may be used.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 46 of 67 Table 8-9 32PHB DSC Pressure Loads Operating Condition Internal Pressure ASME Service Level "Normal Pressure" 15psig A (1% rods ruptured) _5_psigA "Blow Down" 20 psig A "Off-Normal" Pressure 20 psig B (10% rods ruptured) 20_psigB "Accident" Pressure (100% rods ruptured during Transfer Accidents) 100 psig D (10% rods ruptured for blocked vent accident in HSM-HB)

Note:

") Conditions (percentage of rods damaged) are from Section 4.0.V.5.c of NUREG-1536 [2.3.19].

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 47 of 67 Table 8-10 Thermal Conditions for 32PHB DSC Analyses Operating 32PHB DSC Minimum Maximum Solar Ambient Ambient Heat Reference Temperature Temperature Flux Transfer Cask (Fuel Building) (3) (3 N/A [2.4.15]

Transfer Cask -80 F 104 0 F 82 [2.4.4]

Normal Btu/br-ft 82 HSM -8°F 104 0 F [2.4.4]

Btulhr-ft2 - .

Transfer Cask (Fuel Building) (3) (3) N/A ---

127 Off-Normal Transfer Cask -80F 104 0 F Btu/hr1ft2 [2.4.4]

HSM-HB -80F 104 0F 127f2 [2.4.4]

Btu/hr-fi2 [.44 Transfer Cask 127 (Loss of forced air) Btu/hr-f [.

Transfer Cask 127 Accident (Fire Accident) (1)/a 104F Btu/hr-ft 2 [2.4.4]

HSM-HB 127 (Blocked inlet or outlet vents) (2) -80 F 1040 F Btu/hr-fi2 [2.4.4]

Notes:

() The transfer cask fire accident bounds the HSM-HB fire case.

(2) 10% rod rupture is considered for this blocked vent accident condition for DSC internal pressure calculation.

(3) An average ambient temperature of 1007F should be considered within Fuel (Auxiliary) Building

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 48 of 67 Notes:

(1) The design loads listed in this table ensure that the lic ram forces are bounded conservatively.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NULH32PHB-01] 01] REVISION: 4 PROJECT NO: 10955 IPAGE: 49 of 67 9.0 ONSITE TRANSFER CASK STRUCTURAL DESIGN REQUIREMENTS 9.1 Structural Design Criteria The CCNPP-FC TC is designed to meet the criteria of ASME Code, 1992, Subsection NC for Class 2 components. Service Level A and B allowables are used to for all normal operating and off-normal loadings. Service Level C and D allowables are used for load combinations that include postulated accidents loadings.

Cask upper trunnions shall be evaluated for the 32PHB payload in accordance with ANSI N 14.6

[2.1.8]. The local shell stresses at the trunnion locations may be analyzed per Welding Research Council [2.4.28]. Limit analysis specified in Appendix F of the ASME Code,Section III (F.1341.3) can also be used for structural evaluation of the cask for level D events.

The allowable stress criteria for the TC and the TC bolts are summarized in Table 9-10 and Table 9-11, respectively 9.2 Loads and Load Combinations A summary of the load and load combinations used for the 32PHB design was listed in Table 7-2.

This section provides a summary description of the individual loads requiring evaluation for the CCNPP-FC TC.

9.2.1 Deadweight Deadweight loads shall include the self-weight of the transfer cask and the payload of the 32PHB DSC.

9.2.2 Internal Pressure Since the 32PHB DSC provides pressure boundary, this load is not applicable for CCNPP-FC transfer cask.

9.2.3 Thermal The thermal load for the CCNPP-FC TC is 29.6 kW. For the structural analyses, temperature profiles and maximum component temperatures are based on thermal loads listed in Table 8-10, which consider the environmental conditions described in Section 6.0.

The maximum component temperatures shall be considered in determining the allowable stresses for each condition. Maximum thermal gradients are to be considered for determining thermal stresses.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 50 of 67 9.2.4 Seismic The seismic design criteria consists of a maximum horizontal component ground acceleration of

0. 15g and maximum vertical component of 0.1Og as described in Section 3.2.5.3 of the USAR

[2.4.5].

The seismic loads during transfer from the CCNPP Auxiliary Building to the HSM-HB site shall be developed in accordance with NRC Regulatory Guides 1.60 and 1.61 with accelerations of 0.1 5g horizontal and 0.10g vertical, with 3% critical damping. In lieu of a site specific seismic analysis, the DSC may be analyzed for the above seismic accelerations applied as equivalent static loads in accordance with Section 8.2.3.2, Paragraph A.ii, "DSC Seismic Stress Analysis", of the NUTECH Topical Report NUH-002 [2.4.1]. If a dynamic response spectrum analysis for seismic loads is performed, a procedure consistent with NRC Regulatory Guide 1.92 shall be used for combing the response values for individual modes.

9.2.5 Handling Handling loads for CCNPP-FC TC are summarized in Table 8-11.

9.2.6 Drop Loads The accident drop loads, which result from a postulated drop of TC loaded with 32PHB DSC shall be considered. The following drops are to be considered for on-site transfer drops.

Equivalent static deceleration:

75g vertical end drop 75g horizontal side drop 25g corner drop with slap down (corresponds to an 80" drop height)

Structural damping during drop: 10%

The above loads are taken from USAR [2.4.5] for NUHOMS 24P and 32P designs. No evaluation is required for corner drop since the stresses are bounded by the vertical drop stresses as shown in USAR [2.4.5].

9.2.7 Tornado Wind and Missile Loads Tornado wind and missile loads for the HSM are used. Drop loads are normally bounding.

9.2.8 Flood Loads Flood loading is excluded by CCNPP Specification 2.4.4 and need not be considered in the design of the TC.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 I REVISION: 4 ýI PROJECT NO: 10955 PAGE: 51 of 67 Table 9-1 Material Properties - SA-240/SA-479 Type 304 ASTM A-240, Type 304 (TC)

Nominal Composition: 18Cr-8Ni Temp E Sm SY Su 0INST 0XAVG p k a

(°F) (10 ksi) (ksi) (ksi) (ksi) (10-60IF-1) (10-60F-1) (lb/in 3 ) (Btu/hr-fl-°F) (ft2/hr)

-20 20.0 30.0 75.0 70 28.3 8.46 8.46 8.6 0.151 100 20.0 30.0 75.0 8.63 8.55 8.7 0.152 150 8.87 8.67 200 27.6 20.0 25.0 71.0 9.08 8.79 9.3 0.156 250 9.27 8.90 300 27.0 20.0 22.5 66.0 9.46 9.00 9.8 0.160 350 9.64 9.10 400 26.5 18.7 20.7 64.4 9.80 9.19 10.4 0.165 450 9.95 9.28 500 25.8 17.5 19.4 63.5 10.10 9.37 0.290 10.9 0.170 550 10.25 0.45 600 25.3 16.4 18.2 63.5 10.38 9.53 11.3 0.174 650 16.2 17.9 63.5 10.50 9.61 700 24.8 16.0 17.7 63.5 10.60 9.69 11.8 0.179 750 15.6 17.3 63.1 10.70 9.76 800 24.1 15.2 16.8 62.7 10.79 9.82 12.2 0.184 850 16.5 61.9 900 23.5 16.2 61.0 12.7 0.189 950 15.9 59.4 1000 15.6 57.7 13.2 0.194 SA-240, Table Table Table 304 or TM-I Table U A-240, pg 664 2A Y1 pg 494 304 GroupG P9 354 pg 573.1 Table TE-I Perry Calculated based on Table Table TCD Table Table Table pg 640 23-5 pg 656 SA-479, TM-I 2A Y-1 Table U 304 pg 664 pg 358 pg 574 pg 496 Group G ASME Section 1I, Part D - Properties, 1992 Perry Chemical Engineers' Handbook, 5h Edition

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 T REV1ISION: 4_

PROJECT NO: 10955[ PAGE: 52 of 67 Table 9-2 Material Properties - SA 516, Gr. 70 (TC)

Carbon Steel with C < 0.30% (C-Mn-Si)

Temp E Sm Sy Su aINST aAVG P k a (OF) (103 ksi) (ksi) (ksi) (ksi) (10-6 OF-) (10-6OF" 1) (lb/in 3 ) (Btu/hr-fV-°F) (fi2/hr)

-20 23.3 38.0 70.0 70 29.5 5.42 5.42 23.6 0.454 100 23.3 38.0 70.0 5.65 5.53 23.9 0.443 150 6.03 5.71 200 28.8 23.1 34.6 70.0 6.39 5.89 24.4 0.422 250 6.73 6.09 300 28.3 22.5 33.7 70.0 7.04 6.26 24.4 0.406 350 _ 7.33 6.43 400 27.7 21.7 32.6 70.0 7.60 6.61 24.2 0.386 450 7.85 6.77 500 27.3 20.5 30.7 70.0 8.07 6.91 0.284 23.7 0.364 550 8.28 7.06 600 26.7 18.7 28.1 70.0 8.46 7.17 23.1 0.346 650 18.4 27.6 70.0 8.62 7.30 700 25.5 18.3 27.4 70.0 8.75 7.41 22.4 0.320 750 26.5 69.3 8.87 7.50 800 24.2 25.3 64.3 8.96 7.59 21.7 0.298 850 24.4 58.6 900 22.4 24.1 52.0 20.9 0.274 950 23.2 46.2 1000 21.1 40.3 1 20.0 0.248 Table Table Table Table U Table TE-I Calculated based on TM-I 2A Y-1 pg 480 pg 638 AISC Table TCD pg 664 pg 298 pg 514 group C pg 650 ASME Section 11, Part D - Properties, 1992 American Institute of Steel Construction (AISC), 9"' Edition

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 53 of 67 Table 9-3 Material Properties - SA 564, Gr. 630 (TC)

Nominal Composition: 17Cr-4Ni-4Cu Temp E Sm Sy Sy cINST (XAVG P k a (OF) (103 ksi) (ksi) (ksi) (ksi) (10-6 OF-) (10-6°F-l) (lb/in 3 ) (Btu/hr-ft-0 F) (ft 2/hr)

-20 45.0 105.0 135.0 70 28.3 5.89 5.89 9.9 0.188 100 1 45.0 105.0 135.0 5.89 5.89 10.1 0.189 150 5.89 5.89 200 27.6 45.0 97.1 135.0 5.90 5.90 10.6 0.189 250 5.90 5.90 300 27.0 45.0 93.0 135.0 5.90 5.90 11.2 0.190 350 5.91 5.91 400 26.5 43.8 89.8 131.4 5.91 5.91 11.7 0.191 450 5.91 5.91 500 25.8 42.8 87.0 128.4 5.91 5.91 0.284 12.2 0.190 550 5.93 5.93 600 25.3 42.1 84.7 126.7' 5.96 5.93 12.7 0.190 650 41.9 83.6 5.99 5.93 700 24.8 6.03 5.94 13.2 0.186 750 1 6.08 5.95 800 24.1 6.14 5.96 13.5 0.180 850 900 23.5 13.7 0.172 950 1000 13.8 0.160 Table Table Table Table U Table TE- Perry Calculated based on pg 2A Y-1 Table Table TCD pg 664 pg 325.1 pg 546 pg 488 pg 640 23-5 Group G pg 657 ASME Section 11, Part D - Properties, 1992 Perry Chemical Engineers' Handbook, 5th Edition

A TRANSNUCLEAR AN ARE VA COMPANY DOCUMENTNO: NUH32PHB-0101 I REVISION: 4 PROJECT NO: 10955 PAGE: 54 of 67 Table 9-4 Material Properties - SA-182 Type F304N (TC)

Nominal Composition: 18Cr-8Ni-N Temp E Sm SY Su CLINST aAVG p k a (OF) (103 ksi) (ksi) (ksi) (ksi) (10-6 OF-) (10-6 OF-) (lb/in 3 ) (Btu/hr-ft-0 F) (ft 2/hr)

-20 23.3 35.0 80.0 70 28.3 8.46 8.46 8.6 0.151 100 23.3 35.0 80.0 8.63 8.55 8.7 0.152 150 8.87 8.67 200 27.6 23.3 28.7 80.0 9.08 8.79 9.3 0.156 250 9.27 8.90 300 27.0 22.5 25.0 75.9 9.46 9.00 9.8 0.160 350 1 9.64 9.10 400 26.5 20.3 22.5 73.2 9.80 9.19 10.4 0.165 450 9.95 9.28 500 25.8 18.8 20.9 71.2 10.10 9.37 0.290 10.9 0.170 550 10.15 9.45 600 25.3 17.8 19.8 69.7 10.38 9.53 11.3 0.174 650 17.6 19.5 69.1 10.50 9.61 700 24.8 17.2 19.1 68.6 10.60 9.69 11.8 0.179 750 16.9 18.8 68.0 10.70 9.76 800 24.1 16.7 18.5 67.2 10.79 9.82 12.2 0.184 850 18.1 66.4 900 23.5 17.7 65.6 12.7 0.189 950 17.3 64.5 1000 16.8 62.9 13.2 0.194 Table Table Table Table U Table TE- Perry Calculated based on pg 6 2A Y-1 Table Table TCD pg 664 pg 358 pg 578 pg 496 pg 640 23-5 group G pg 656 ASME Section II, Part D - Properties, 1992 Perry Chemical Engineers' Handbook, 5th Edition

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB-0101 I REVISION: 4 PROJECT NO: 10955 1 PAGE: 55 of 67 Table 9-5 Material Properties - SA-193 Gr. B7 (TC)

Temp Maximum Allowable Stress Value Yield Strength, Sy Temp for Class 2 Component, S (kYS)

(ksi) (ksi)

-20 25.0 105.0 100 25.0 105.0 150 25.0 200 25.0 98.0 300 25.0 94.1 400 25.0 91.5 500 25.0 88.5 550 25.0 600 25.0 85.3 650 25.0 83.0 700 25.0 80.6 750 23.6 77.5 800 21.0 74.0 850 17.0 900 12.5 950 8.5 Table 3 Table pg 424 Y1 pg 530 ASME Section H, Part D - Properties, 1992

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: _4 PROJECT NO: 10955l PAGE: 56 of 67 I Table 9-6 Mechanical Properties for ASTM B29 Lead (DSC and TC)

(Static Properties)

Static Stress Propertie (ksi) E Coef. Of Temp. Yield (Sy) Ultimate (Su) (106 psi) Thermal Exp

(°F) Tension Compression Tension (10-6 in/in/°F)

-99 - - 2.50 15.28 70 - - - 2.34 16.07 100 0.584 0.490 1.570 2.30 16.21 175 0.509 0.428 1.162 2.20 16.58 250 0.498 0.391 0.844 2.09 16.95 325 0.311 0.320 0.642 1.96 17.54 440 - - - 1.74 18.50 620 - - 1.36 20.39 Note: The material properties of the lead are based on "Safety Analysis Report for the NUHOMS - MP187 Multi-Purpose Cask", Docket 71-9255, TN Document Number NUH-05-15 1, Revision 17, July 2003 (Table 2.3.4-2).

(Dynamic Stress-Strain Properties)

Strain Stress (ksi)

(in/in) 100°F 230TF 300°F 350OF 500°F 0.000485 1.14 1.06 1.00 0.97 0.86 0.03 2.2 2.0 1.7 1.5 1.1 0.1 3.3 2.8 2.38 2.1 1.26 0.3 4.9 3.2 2.72 2.4 1.44 0.5 5.6 3.6 3.06 2.7 1.62 Note: The static and dynamic stress properties of the lead are taken from Tietz [2.4.29]. The Young's modulus and coefficient of thermal expansion values are taken from NUREG/CR-0481[2.3.24], Pages 56 and 66.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 I PROJECT NO: 10955 PAGE: 57 of 67 Table 9-7 Thermal Properties for Lead (DSC and TC)

Thermal Properties Temp Thermal Specific Heat Density Conductivity SpecficHeaDesit (K) (W/m-K) (kJ/kg-K) .(kg/mr3) 100 39.7 0.118 11,520 150 37.9 0.122 11,470 200 36.7 0.125 11,430 250 36.0 0.127 11,380 300 35.3 0.129 11,330 400 34.0 0.132 11,230 500 32.8 0.137 11,130 600 31.4 0.142 11,010 Ch. 3, Table68 Ch. 3, Table 67 Ch. 3, Table 64

Reference:

Rohsenow, W.M., "Handbook of Heat Transfer", 2 nd Edition, McGraw-Hill Handbooks.

Table 9-8 Mechanical and Thermal Properties for NS-3 (TC)

Mechanical Properties Poisson Ratio Compressive Elasticity Strength (ksi) (1.0E3 ksi) 0.2 3.9 0.16 Thermal Properties

Reference:

Calculations 1095-35, Rev. 2 and Calculation 1095-5, Rev. 0.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 58 of 67 Table 9-9 Material Properties - Air (DSC and TC)

Temperature Thermal conductivity Temperature Thermal conductivity (K) (W/m-K) (OF) (Btu/hr-in-°F) 200 0.01822 -100 0.0009 250 0.02228 -10 0.0011 300 0.02607 80 0.0013 400 0.03304 260 0.0016 500 0.03948 440 0.0019 600 0.04557 620 0.0022 800 0.05698 980 0.0027 1000 0.06721 1340 0.0032 The above data are calculated based on the following polynomial function from: Rohsenow, W.M., "Handbook of Heat Transfer", 3rd Edition, McGraw-Hill Handbooks.

k = 2 Ci T' for conductivity in(W/m-K) and T in (K)

For 250 < T < 1050 K CO -2.27650!0E-03 C1 1.2598485E-04 C2 -1.4815235E-07 C3 1.7355064E-10 C4 -1.0666570E-13 C5 2.4766304E-17 Specific heat, viscosity, density and Prandtl number of air are used to calculate heat transfer coefficients based on the following data from: Rohsenow, W.M., "Handbook of Heat Transfer", 3rd Edition, McGraw-Hill Handbooks.

S= J A T' for specific heat in (kJ/kg-K) and T in (K)

For 250 < T < 1050 K AO 0.103409E+1 Al -0.2848870E-3 A2 0.7816818E-6 A3 -0.4970786E-9 A4 0.1077024E-12 t=-Bi T' for viscosity (N-s/m 2)xl06 and T in (K)

For 250 < T < 600 K For 600 < T < 1050 K BO -9.8601 E- 1 BO 4.8856745 BI 9.080125E-2 BI 5.43232E-2 B2 -1.17635575E-4 B2 -2.4261775E-5 B3 1.2349703E-7 B3 7.9306E-9 B4 -5.7971299E-I1 B4 -1.10398E-12 p = PIRT for density (kg/m3) with P=101.3 kPa; R = 0.287040 kJ/kg-K; T = air temp in (K);

Pr = cp pl/k Prandtl number

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PfIB-0101 FREVISION:- 4 I

PROJECT NO: 10955 PAGE: 59 of 67 Table 9-10 Structural Stress Criteria for Transfer Cask Stress Values Item Stress Type Service Levels Service Level Service Level D Service Level D A& B C (Elastic Analysis) (Plastic Analysis)

Larger of 0.7 Su or Primary Membrane Sm 1.2 Sm or 0.7Sru m Sy + (S -SO Transfer Cask 3 Transfer CaskSmaller of 3.6 Sm 09S Structural Primary Membrane 1.5 Sm 1.8 Sm or Shell + Bending 1.5.8_or Su_0 Primary + N/A N/A N/A Secondary Membrane and Smaller of Membrane + 5 /6 or 5 10 N/A N/A N/A Trunnions(l) Bending Sy/6_orS_/10 Smaller of Shear N/A N/A N/A S__6 orS/10 N/

Smaller of 1.2 SmN/

Primary 0.5Sm 0.6 Sm Salro1.S.N/A Fillet Welds or 0.35 Su Secondary 0.75 Sm N/A N/A N/A Notes:

(1) These allowables apply to the upper lifting trunnions for cask vertical lifts within the Auxiliary/Fuel Building. The lower support trunnions and the upper lifting trunnions for all remaining loads are governed by the same ASME Code criteria applied to the cask structural shell.

Table 9-11 Structural Stress Criteria for Transfer Cask Bolts Service Levels A, B, and C Average Service Stress < 2 Sm Maximum Service Stress < 3 Sm Service Level D Average Tension Smaller of Sy or 0.7 Su Tension + Bending Su Average Shear Smaller of 0.6 Sy or 0.42 Su Interaction equation of Appendix F (F- 1335.3) of ASME Code [2.1.4]

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 60 of 67 10.0 HSM-HB STRUCTURAL DESIGN CRITERIA The HSM-HB design criteria are the same as those utilized for the HSM-H with the 24PTH/32PTH DSCs and documented in CoC 1004 UFSAR, Appendix P of [2.4.4] and in CoC 1030 SAR [2.4.13].

11.0 THERMAL REQUIREMENTS Thermal properties of materials including material temperature limits should be traced to authoritative references. For materials not listed in appropriate references, thermal properties shall be obtained by testing or from other verifiable sources.

Thermal analysis for the 32PHB shall be based on fuel assemblies with decay heat up to 29.6 kW per canister. Zoning is used to accommodate per assembly heat loads as high as 1.0 kW as shown in Figure 4-1. The DSC system shall be passively cooled as required in 10 CFR 72.236(0.

Forced air circulation may be required for transfer operation. Time to initiate air circulation following drainage of the TC/DSC annulus shall be greater than or equal to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months for the maximum heat load.

In addition to evaluation of the design basis heat load, the design shall also identify maximum heat loads with uniform heat load zoning configurations, which allow forced cooling initiation times of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Peak clad temperature of the fuel at the beginning of the long term storage shall not exceed 400'C for normal conditions of storage, and for short-term operations, including canister drying and backfilling, according to ISG I1, Revision 3 [2.3.23].

Fuel cladding (Zircaloy) temperature shall be maintained below 570'C (1058'F) [2.3.23] for off-normal and accident conditions.

For DSC unloading operations, cladding integrity should be maintained during reflooding, so as not to interfere with fuel handling and retrieval.

Decay heat shall be calculated with the radiological source terms.

Insolation is 82 Btu/hr-ft2 for normal conditions and 127 Btu/hr-ft2 for off-normal and accident conditions based on [2.4.4], Section 3.5.1.

Fuel cladding and basket material temperatures should be calculated assuming steady state conditions during vacuum drying operations. If calculated temperatures are not acceptable, transient analysis should be performed assuming limited time period for vacuum drying operations. The blowdown during vacuum drying operation may be performed using helium or nitrogen.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 61 of 67 Based on ISG-I 1, Rev. 3 [2.3.23], repeated thermal cycling of the cladding during fuel loading operations is limited to ten cycles and the thermal cycling of the cladding with temperature reductions greater than 650 C is not permitted.

Transfer Cask, HSM-HB, DSC, and Fuel Cladding materials shall be maintained within their minimum and maximum temperature criteria for normal, off-normal and accident conditions.

Assume for analysis purposes that all HSM-HB inlet or all outlet vents are blocked.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101] REVISION: 4 PROJECT NO: 10955 PAGE: 62 of 67 12.0 SHIELDING REQUIREMENTS The design basis neutron and gamma source terms are provided by Constellation Energy, Nuclear Generation Group and shown in Table 4-4 through Table 4-6. Upholding the restrictions on the number of the reconstituted fuel rods and the requirements for additional cooling time specified in

[2.4.30] and [2.4.31], the source terms presented in Table 4-4 through Table 4-6 are applicable for reconstituted fuel assemblies.

The 32PHB DSC shall be designed to limit radiation exposure to both operators and the general public in accordance with ALARA.

For storage the radiation shielding must meet the requirements of 10CFR72.104 and I OCFR72.106.

Based on [2.4.4], Section 3.6, the dose rates shall be limited to the following values:

Contact dose rate on the exterior surface of the transfer cask _<250 mrem/hr

" Contact dose rate on the dxterior surface of the HSM-HB shield door < 100 mrem/hr

" Contact dose rate on the exterior surface of the HSM-HB sides and roof, excluding the vents

< 20 mrem/hr After a design basis accident an individual at the boundary or outside the controlled area shall not receive a dose rate greater than 5 rem to the whole body or to any organ.

Doses calculated for workers and the public shall comply with the criteria in 10 CFR 20 and 72.

Gammas with energies from approximately 0.8 to 2.5 Mev will be considered as significant contributors to the dose rate.

The contribution from the irradiated fuel assembly hardware to the source term and the dose rate shall also be considered.

The flux-to-dose rate conversion factor shall be based on ANSI/ANS 6.1.1-1977.

Degradation of shielding material at higher temperature, if applicable, shall be accounted for in the shielding evaluation.

A TRANSNUCLEAR AN AREVA COMPANY IDOCUMENT NO: NUH32PHB-0101 J.REVISION: 4_

PROJECT NO: 10955 PAGE: 63 of 67 13.0 CRITICALITY REQUIREMENTS The criticality analysis shall determine the minimum poison loading requirements as a function of fuel enrichment. The loading requirements are the thickness and minimum absorber loading (B10) utilized in the criticality analysis and the type of poison, if the analysis establishes a statistically significant variation in the system reactivity due to poison type. The credit for the amount of the absorber material in the criticality analysis shall be detailed in the material specification based on the poison material like, borated aluminum alloy or metal matrix composites (MMCs) following requirements of NUREG CR-5661 [2.3.21]. Poison plates may be composed entirely of borated material or may be thinner borated sheet paired with aluminum sheet to achieve the required thickness.

13.1 General Criticality Criteria No credit for burnable poison materials within the fuel assemblies as a neutron absorber shall be taken.

For a single DSC or an array of DSCs, the effective criticality factor, ker, shall not exceed 0.95 with a 95% probability at a 95% confidence level including uncertainties under all credible normal, off-normal, and accident conditions. Model bias and benchmarking bias shall be accounted for in the criticality analysis.

Assume no more than 90% of the poison material is effective for MMC and borated aluminum. To allow 90% credit poison material coupons shall be tested via neutron transmission plus radiography.

Criticality analysis shall consider reconstituted fuel that replaces same amount of water as the original fuel pins.

The canister shall be designed and fabricated such that the spent fuel is maintained in a subcritical condition under all credible normal, off-normal, and accident conditions (10 CFR 72.124(a) and 72.23(c)).

The criticality analysis shall demonstrate that the fuel assembly used as the design basis is the most reactive. For the fuel assemblies with axial variation in enrichments, criticality analysis shall demonstrate that the maximum enrichment selected for the fuel assemblies is bounding.

The criticality analysis needs to consider bounding enrichments from the values given in Table 4-2.

The criticality analysis must demonstrate that the DSC remains subcritical for all credible conditions of moderation.

Criticality control shall be demonstrated with a combination of fixed geometry, neutron poison in the basket, and soluble Boron in the pool, as appropriate. Credit for soluble boron shall be limited to 2600 ppm.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENTNO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 IPAGE: 64 of 67 To ensure the compliance with the criticality acceptance criteria of kfr < 0.95 as delineated in ANSI/ANS 57.2 - 1983, full and optimum moderator density conditions over 0.1 to 1.0 g/cc range shall be considered during wet loading and unloading of the fuel assemblies.

The fuel assembly misloading will not be considered for criticality evaluation.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: N-UH32PHB-01]01 ]7EýV IýSION: _4 PROJECT NO: 10955 1 PAGE: 65 of 67 14.0 CONFINEMENT/CONTAINMENT CRITERIA The canister must maintain confinement of radioactive material within the limits of 10 CFR 72.236(l) and 10 CFR 20 under normal, off-normal, and credible accident conditions.

Canister shall be designed and tested to meet the leak-tight criteria defined in ISG-18 [2.3.23] and ANSI N 14.5, Ref. [2.1.8].

15.0 ACCEPTANCE TESTING Testing for the 32PHB DSC shall include those required by the ASME Code for the qualification of materials, welded joints, and canister leakage per ANSI N14.5-1997. Additionally, specific operational type fit-up testing is required. Specific tests required shall be specified in the appropriate drawings or included in the canister fabrication specification.

The following minimum testing shall be performed:

ASME Code required testing (materials and welding),

Helium leak testing of the final pressure boundary to a "leak tight" condition as defined by ANSI N 14.5 -1997,

Dummy fuel assembly insertion and withdrawal for each basket fuel compartment, and,

" Testing of the poison matrix material. These tests shall provide the justification for assuming that 90% of the poison material is effective for metal matrix composite and borated aluminum.

DSC shell assembly closure welds shall be in accordance with the guidance of NRC's Interim Staff Guidance ISG- 15 [2.3.23]. Other 32PHB DSC basket welds shall identify efficiency factors and inspection criteria in the calculation.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 REVISION: 4 PROJECT NO: 10955 PAGE: 66 of 67 16.0 MATERIAL REQUIREMENTS 16.1 Specifications Materials meeting the requirements of ASME B&PV Code,Section III, Article NB-2000, and the specification requirements of Section II, shall be used in the design to the maximum extent practical.

Detailed procurement specifications shall be required for other materials to assure that mechanical and other property values used in the design calculations will be met.

16.2 Properties The material properties, stress intensity values and allowable stresses shall be obtained from the ASME B&PV Code,Section II, Part D.

For other materials, the source of material property data shall be identified and documented.

Materials shall be selected based on their corrosion resistance, susceptibility to stress corrosion cracking, embrittlement properties, and the environment in which they operate during normal and accident conditions. The lowest service temperature for metallic component in the DSC is -80 F.

16.3 Impact Properties Test The 32PHB DSC components shall be evaluated for their impact properties and shall meet the requirements of the applicable material specifications (ASME B&PV Code, Section 11) and ASME B&PV Code Section III, Subsection NB-2000 and Subsection NG-2000.

16.4 Materials Suitability (Chemical, Galvanic and Other Reactions)

Materials suitability shall be reviewed in accordance with 10 CFR 72, NRC Bulletin 96-04 and IOCFR71.44 (d). Materials and construction shall be selected to assure that there will be no significant chemical, galvanic, or other reaction among packaging components and contents.

Materials shall be chosen that will preclude a galvanic effect that could lead to unacceptable fuel cladding corrosion or generate flammable gases in unacceptable quantities. Combustible gases should not exceed 2.4% of free gas volume in any confined region for both normal and hypothetical accident conditions [2.4.21 ].

Material suitability evaluation should include:

the possible reaction from water inleakage,

the behavior of materials under irradiation, and

e.

A TRANSNUCLEAR AN AREVA COMPANY DOCUMENT NO: NUH32PHB-0101 7REVISION: 4 PROJECT NO: 10955l PAGE: 67 of 67 the behavior materials during all operations, e.g. operating temperatures and loading pool environment.

16.5 Protective Coatings The materials used for protective coatings (if required) shall be compatible with the DSC materials, operating temperatures, loading pool environment and other interfacing materials or components.

16.6 Emissivities Emissivity values for various surfaces important for heat transfer shall be specified in the calculations. The stainless steel emissivity in calculation of the effective fuel conductivity shall remain between 0.3 and 0.35.

16.7 Effects of Radiation Construction materials shall be compatible with the expected radiation levels.

16.8 Prohibited / Hazardous Materials The design shall not include sulfur, mercury, asbestos, low melting point metals, their alloys or components.

Materials in contact with pool water shall not release materials that contain chlorine or other halogens, sulfur, nitrates, mercury, lead, zinc, copper, tin, gallium, arsenic, antimony, bismuth, silver, cadmium or indium.

17.0 QUALITY ASSURANCE REQUIREMENTS The Important to Safety components of the NUHOMS 32PHB DSC shall be designed, in accordance with the most recent revision of Transnuclear's Quality Assurance Program.

ML15075A339

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SUMMARY

SUMMARY

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