Text

Tn Americas LLC, Consolidated Safety Analysis Report Associated with the Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No. TN-40 Packaging
	ML23341A013
Person / Time
Site:	07109313
Issue date:	12/06/2023
From:	Shaw D; TN Americas LLC
To:	; Office of Nuclear Material Safety and Safeguards, Document Control Desk
Shared Package
ML23341A012	List: ML23341A013;
References
	E-62614
	Download: ML23341A013 (1)
	v • d • e

December 6, 2023 E-62614 Orano TN 7160 Riverwood Drive Suite 200 U. S. Nuclear Regulatory Commission Columbia, MD 21046 Attn: Document Control Desk USA One White Flint North Tel: 410-910-6900 Fax: 434-260-8480 11555 Rockville Pike Rockville, MD 20852

Subject:

Consolidated Safety Analysis Report Associated with the Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No.

TN-40 Packaging, Docket No. 71-9313

References:

(1) TN Letter E-59049, Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No. TN-40 Packaging, Docket No.

71-9313, dated December 16, 2021 (Agencywide Documents Access and Management System [ADAMS] Package Accession No. ML21350A282)

(2) TN Letter E-62022, Response to Request for Additional Information for the Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No. TN-40 Packaging, Docket No.

71-9313, dated June 6, 2023 (ADAMS Package Accession Nos.

ML23157A029, ML23157A030)

(3) TN Letter E-62829, Supplemental Information for the Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No.

TN-40 Packaging, Docket No. 71-9313, dated November 14, 2023 (ADAMS Package Accession No. ML23318A179)

(4) TN Letter E-62984, Supplemental Information for the Application for Revision 4 to Certificate of Compliance No. 9313 for the Model No.

TN-40 Packaging, Docket No. 71-9313, dated December 5, 2023 (ADAMS Package Accession Nos. ML23339A062, ML23339A063, ML23339A064)

In accordance with 10 CFR 71.31, TN Americas LLC (TN Americas) submitted an application for Revision 4 to Certificate of Compliance (CoC) No. 9313 for the TN-40 packaging [1], as supplemented to provide additional changes [2], and as supplemented to provide additional information requested by the NRC to complete the review in accordance with 10 CFR 71.38 [3 and 4]. The purpose of this submission is to provide a consolidated TN-40 Safety Analysis Report (SAR), Revision 17 to reference as the application for CoC No. 9313, Revision 4.

Enclosures transmitted herein contain SUNSI. When separated from enclosures, this transmittal document is decontrolled.

E-62614 Document Control Desk Page 2 of 2 The consolidated TN-40 SAR Revision 17 is included as Enclosure 1. A public version of the Revision 17 SAR with proprietary information redacted is provided for public availability as . In accordance with 10 CFR 2.390, TN Americas is providing an affidavit as , requesting that this proprietary information be withheld from public disclosure.

Should the NRC staff require additional information to support review of this application, please contact Peter Vescovi at 336-420-8325 or peter.vescovi@orano.group.

Sincerely, Digitally signed by SHAW SHAW Donis Date: 2023.12.06 Donis 11:25:17 -05'00' Don Shaw Licensing Manager cc: Pierre Saverot (NRC), Senior Project Manager, Storage and Transportation Licensing Branch, Division of Fuel Management Scott Bomar, Project Manager, TN Americas

Enclosures:

1. TN-40 Safety Analysis Report, Revision 17 (Proprietary Version)

2. TN-40 Safety Analysis Report, Revision 17 (Public Version)

3. Affidavit Pursuant to 10 CFR 2.390

Enclosure 1 to E-62614 TN-40 Safety Analysis Report, Revision 17 (Proprietary Version)

Withheld Pursuant to 10 CFR 2.390

Enclosure 2 to E-62614 TN-40 Safety Analysis Report, Revision 17 (Public Version)

PUBLIC TN-40 Transportation Packaging Safety Analysis Report Prepared by:

TN Americas LLC Columbia, Maryland Revision 17 November 2023

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 RECORD OF REVISIONS E-Document Transmittal Letter Revision Number Date E-58661 April 12, 2021 E-59049 December 16, 2021 17 E-62829 November 14, 2023 E-62984 December 5, 2023 E-62614 December 6, 2023 Page i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 TABLE OF CONTENTS Introduction .......................................................................................... 1-1 Package Description ............................................................................ 1-1 Packaging ............................................................................... 1-1 Operational Features .............................................................. 1-7 Contents of Packaging ............................................................ 1-8 References ......................................................................................... 1-10 Appendices ........................................................................................ 1-11 TN-40 Packaging Drawings .................................................. 1-11 Structural Design.................................................................................. 2-1 Discussion .............................................................................. 2-1 Design Criteria ........................................................................ 2-3 Weights and Center-Of-Gravity ............................................................ 2-6 Mechanical Properties of Materials ...................................................... 2-6 Cask Material Properties ........................................................ 2-6 Basket Material Properties ...................................................... 2-7 Impact Limiter Material Properties .......................................... 2-7 Fracture Toughness Requirements ........................................ 2-7 General Standards For All Packages ................................................... 2-7 Minimum Package Size .......................................................... 2-7 Tamper-proof Feature ............................................................. 2-7 Positive Closure ...................................................................... 2-7 Chemical and Galvanic Reactions .......................................... 2-8 Lifting And Tie-Down Standards......................................................... 2-10 Lifting Devices ...................................................................... 2-10 Tie-Down Devices ................................................................. 2-16 Normal Conditions Of Transport......................................................... 2-17 Heat ...................................................................................... 2-18 Cold Environment ................................................................. 2-19 Increased External Pressure (N4)......................................... 2-20 Reduced External Pressure (N5) .......................................... 2-20 Transport Shock Loading (N14 & N15) ................................. 2-20 Page TOC-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Transport Vibration Loading (N12 & N13) ............................. 2-21 Water Spray .......................................................................... 2-21 Free Drop (N6 through N11) ................................................. 2-21 Corner Drop .......................................................................... 2-22 Compression......................................................................... 2-22 Penetration ........................................................................... 2-22 Lid Bolt Analysis ................................................................... 2-22 Fatigue Analysis of the Containment Boundary .................... 2-22 Structural Evaluation of the Basket under Normal Condition Loads .................................................................... 2-26 Summary of NCT Cask Body Structural Analysis ................. 2-26 Hypothetical Accident Conditions ....................................................... 2-27 30 Foot Free Drop ................................................................ 2-28 Puncture ............................................................................... 2-33 Thermal ................................................................................ 2-36 Water Immersion .................................................................. 2-37 Structural Evaluation of the Basket under Accident Loads .................................................................................... 2-39 Summary of HAC Cask Body Structural Analysis ................. 2-40 Special Forms / Fuel Rods ................................................................. 2-40 Special Form......................................................................... 2-40 Fuel Rods ............................................................................. 2-40 References ......................................................................................... 2-41 Appendices ........................................................................................ 2-43 ASME Code Alternatives .................................................................... 2-44 2.10.1.1 Introduction ............................................................................. 2.10.1-1 2.10.1.2 ANSYS Analysis ..................................................................... 2.10.1-2 2.10.1.3 ANSYS Analysis Results and Report Methodology .............. 2.10.1-22 2.10.1.4 Trunnion Local Stress Analysis Due To Lifting Load ............ 2.10.1-23 2.10.1.5 References ........................................................................... 2.10.1-24 2.10.2.1 Introduction ............................................................................. 2.10.2-1 2.10.2.2 Lid Bolt Load Calculations ...................................................... 2.10.2-2 2.10.2.3 Summary of Lid Bolt Loads ..................................................... 2.10.2-8 Page TOC-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.4 Lid Bolt Load Combinations .................................................... 2.10.2-9 2.10.2.5 Lid Bolt Stress Calculations .................................................. 2.10.2-11 2.10.2.6 Analysis Results ................................................................... 2.10.2-14 2.10.2.7 Lid Bolt Fatigue Analysis....................................................... 2.10.2-14 2.10.2.8 Lid Seal Contact Evaluation .................................................. 2.10.2-18 2.10.2.9 Minimum Engagement Length for Bolt and Flange ............... 2.10.2-19 2.10.2.10 Conclusions .......................................................................... 2.10.2-21 2.10.2.11 References ........................................................................... 2.10.2-22 2.10.3.1 Introduction ............................................................................. 2.10.3-1 2.10.3.2 Description .............................................................................. 2.10.3-1 2.10.3.3 Materials Input Data ................................................................ 2.10.3-1 2.10.3.4 Applied Loads ......................................................................... 2.10.3-1 2.10.3.5 Method of Analysis ................................................................. 2.10.3-2 2.10.3.6 Analysis Results ..................................................................... 2.10.3-4 2.10.3.7 References ............................................................................. 2.10.3-6 2.10.4.1 Introduction ............................................................................. 2.10.4-1 2.10.4.2 Fracture Toughness Requirements of The Cask .................... 2.10.4-1 2.10.4.3 Fracture Toughness Evaluation of Cask Components and Welds ...................................................................................... 2.10.4-3 2.10.4.4 Methodology ........................................................................... 2.10.4-3 2.10.4.5 Loadings ................................................................................. 2.10.4-3 2.10.4.6 Material Fracture Toughness .................................................. 2.10.4-3 2.10.4.7 Fracture Toughness Criteria ................................................... 2.10.4-5 2.10.4.8 Stress Intensity Factor Calculations ........................................ 2.10.4-5 2.10.4.9 Conclusions ............................................................................ 2.10.4-5 2.10.4.10 NDE Inspection Plan ............................................................... 2.10.4-6 2.10.4.11 References ............................................................................. 2.10.4-7 2.10.5.1 Introduction ............................................................................. 2.10.5-1 Page TOC-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.5.2 TN-40 Fuel Basket Stress Analysis ........................................ 2.10.5-3 2.10.5.3 TN-40 Fuel Basket Buckling Analysis ................................... 2.10.5-11 2.10.5.4 Fusion Welds ........................................................................ 2.10.5-14 2.10.5.5 Sensitivity Analysis ............................................................... 2.10.5-15 2.10.5.6 Summary .............................................................................. 2.10.5-19 2.10.5.7 References ........................................................................... 2.10.5-24 2.10.6.1 Introduction ............................................................................. 2.10.6-1 2.10.6.2 Modal Analysis of Basket Side Drop Loading Condition ......... 2.10.6-1 2.10.6.3 Frequency of Basket due to Basket End Drop Loading Condition................................................................................. 2.10.6-3 2.10.6.4 Dynamic Load Factor Calculations ......................................... 2.10.6-3 2.10.6.5 References ............................................................................. 2.10.6-5 2.10.7.1 Side Drop Analysis ................................................................. 2.10.7-1 2.10.7.2 End Drop Analysis .................................................................. 2.10.7-3 2.10.7.3 Material Properties of Fuel .................................................... 2.10.7-12 2.10.7.4 Determination of Side Drop g-Loading with Dynamic Load Factor ......................................................................................... 2.10.7-12 2.10.7.5 References ........................................................................... 2.10.7-14 2.10.8.1 Introduction ............................................................................. 2.10.8-1 2.10.8.2 Design Description .................................................................. 2.10.8-1 2.10.8.3 Design Criteria ........................................................................ 2.10.8-3 2.10.8.4 Analysis of the HAC 30 Foot Free Drop .................................. 2.10.8-4 2.10.8.5 Analysis for One Foot Drop Normal Condition of Transport .... 2.10.8-7 2.10.8.6 Impact Limiter Attachment Analysis ........................................ 2.10.8-8 2.10.8.7 Summary of ADOC Results Used for Structural Analysis ..... 2.10.8-20 2.10.8.8 Summary Description of ADOC Computer Code .................. 2.10.8-21 2.10.8.9 References ........................................................................... 2.10.8-32 Page TOC-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.9.1 Introduction ............................................................................. 2.10.9-1 2.10.9.2 Scaling Relationships ............................................................. 2.10.9-1 2.10.9.3 Test Article Description ........................................................... 2.10.9-2 2.10.9.4 Test Description ...................................................................... 2.10.9-3 2.10.9.5 Data Measurement ................................................................. 2.10.9-7 2.10.9.6 Test Data and Results ............................................................ 2.10.9-8 2.10.9.7 Conclusions .......................................................................... 2.10.9-16 2.10.9.8 References ........................................................................... 2.10.9-17 2.10.10.1 Hoop Compressive Stress and Buckling Stresses ................ 2.10.10-1 2.10.10.2 Axial Compressive Stress and Buckling Stresses................. 2.10.10-2 2.10.10.3 Amplified Axial Stress ........................................................... 2.10.10-3 2.10.10.4 Interaction Equation for Local Buckling ................................. 2.10.10-4 2.10.10.5 References ........................................................................... 2.10.10-4 2.10.11.1 Introduction ........................................................................... 2.10.11-1 2.10.11.2 TN-40 Impact Limiter Benchmark Analysis ........................... 2.10.11-1 2.10.11.3 TN-40 Lid Closure Evaluation Due to Delayed Impact .......... 2.10.11-4 2.10.11.4 Results of Lid Closure Evaluation Due to Delayed Impact .... 2.10.11-7 2.10.11.5 Conclusion ............................................................................ 2.10.11-9 2.10.11.6 References ......................................................................... 2.10.11-10 3.1 Discussion ............................................................................................ 3-1 3.2 Summary of Thermal Properties of Materials ....................................... 3-3 3.3 Technical Specifications for Components ............................................ 3-8 3.4 Thermal Evaluation for Normal Conditions of Transport ...................... 3-8 3.4.1 Thermal Models ...................................................................... 3-8 3.4.2 Maximum Temperatures ....................................................... 3-14 3.4.3 Maximum Accessible Surface Temperature in the Shade ................................................................................... 3-14 3.4.4 Minimum Temperatures ........................................................ 3-16 Page TOC-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.4.5 Maximum Internal Pressure .................................................. 3-16 3.4.6 Maximum Thermal Stresses ................................................. 3-16 3.4.7 Evaluation of Cask Performance for Normal Conditions of Transport ........................................................ 3-16 3.5 Thermal Evaluation for Hypothetical Accident Conditions .................. 3-19 3.5.1 Fire Accident Evaluation ....................................................... 3-19 3.5.2 Boundary Conditions for the HAC ......................................... 3-19 3.5.3 Crushed Impact Limiter Models ............................................ 3-20 3.5.4 Summary of Results ............................................................. 3-21 3.5.5 Evaluation of Package Performance During and after the HAC Fire ......................................................................... 3-21 3.6 References ......................................................................................... 3-22 3.7 Appendices ........................................................................................ 3-24 3.7.1.1 Discussion ................................................................................ 3.7.1-1 3.7.1.2 14 x 14 PWR Fuel Geometry Parameters................................. 3.7.1-1 3.7.1.3 Summary of Material Properties ............................................... 3.7.1-1 3.7.1.4 Thermal Model .......................................................................... 3.7.1-3 3.7.1.5 Effective Density and Specific Heat .......................................... 3.7.1-9 3.7.1.6 Conclusion .............................................................................. 3.7.1-10 3.7.1.7 References ............................................................................. 3.7.1-11 3.7.2.1 Discussion ................................................................................ 3.7.2-1 3.7.2.2 References ............................................................................... 3.7.2-2 3.7.3.1 Discussion ................................................................................ 3.7.3-1 3.7.3.2 Conclusion ................................................................................ 3.7.3-2 3.7.3.3 References ............................................................................... 3.7.3-2 Containment Boundary ........................................................................ 4-1 Containment Vessel ................................................................ 4-1 Page TOC-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Containment Penetrations ...................................................... 4-1 Seals and Welds ..................................................................... 4-2 Closure ................................................................................... 4-2 Requirements For Normal Conditions Of Transport ............................. 4-3 Containment of Radioactive Material ...................................... 4-3 Pressurization of Containment Vessel .................................... 4-8 Containment Criterion ........................................................... 4-10 Containment Requirements for Hypothetical Accident Conditions .......................................................................................... 4-10 Source Terms ....................................................................... 4-11 Containment of Radioactive Material .................................... 4-11 Containment Criterion ........................................................... 4-11 Leakage Rate Tests for Type B Packages ......................................... 4-13 Special Requirements ........................................................................ 4-14 References ......................................................................................... 4-14 Discussion and Results ........................................................................ 5-1 Source Specification ............................................................................ 5-3 Axial Source Distribution ......................................................... 5-4 Gamma Source....................................................................... 5-5 Neutron Source....................................................................... 5-6 Source Conversion Factors .................................................... 5-6 Fuel Qualification .................................................................... 5-7 Model Specification ............................................................................ 5-10 Description of Radial and Axial Shielding Configuration ........................................................................ 5-11 Shield Regional Densities ..................................................... 5-12 Shielding Evaluation........................................................................... 5-12 Uncertainties and Conservatism in the Shielding Evaluation ............. 5-15 References ......................................................................................... 5-17 Input File Listing ................................................................................. 5-18 SAS2H/ORIGEN-S Input File................................................ 5-18 MCNP Neutron Model Input File ........................................... 5-19 MCNP Primary Gamma Input File ........................................ 5-38 Page TOC-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Discussion and Results ........................................................................ 6-1 Package Fuel Loading ......................................................................... 6-3 Model Specification .............................................................................. 6-4 Description of Calculational Model.......................................... 6-4 Package Regional Densities ................................................. 6-13 Criticality Calculations ........................................................................ 6-19 Calculational Method ............................................................ 6-20 Scoping Calculations and Results ........................................ 6-23 Criticality Calculations and Results ....................................... 6-36 Sensitivity of keff to Number of Isotopes for Burnup Credit .................................................................................... 6-38 Sensitivity of keff to Am-241 ................................................... 6-39 Sensitivity of keff to Gd-155 ................................................... 6-39 Sensitivity of keff to BE Ratio ................................................. 6-39 Sensitivity of keff to Specific Power........................................ 6-40 Summary of Sensitivity Evaluations ...................................... 6-40 Critical Benchmark Experiments ........................................................ 6-40 Benchmark Experiments and Applicability ............................ 6-42 Results of the Benchmark Calculations ................................ 6-44 References ......................................................................................... 6-46 Input File Listing ................................................................................. 6-48 SAS2H Input Deck for Design Basis Fuel Assembly -

Zone 8 .................................................................................. 6-48 CSAS25 Input Deck for Design Basis Criticality Case .......... 6-50 Loading the Wrong Fuel Assembly ....................................................6A-1 6A.1.1 Summary of Preventive Measures ........................................6A-2 Calculating a Burnup Value Higher Than Actual ................................6A-2 6A.2.1 Summary of Preventive Measures ........................................6A-3 Wrong Burnup Value Assigned to a Fuel Assembly ...........................6A-3 6A.3.1 Summary of Preventive Measures ........................................6A-5 Conclusion .........................................................................................6A-5 Page TOC-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 7.1 Package Loading ................................................................................. 7-1 7.1.1 Preparation for Loading .......................................................... 7-1 7.1.2 Loading of Contents................................................................ 7-2 7.1.3 Preparation for Transport ........................................................ 7-2 7.2 Package Unloading .............................................................................. 7-4 7.2.1 Receipt of Package from Carrier............................................. 7-4 7.2.2 Preparation for Unloading ....................................................... 7-5 7.2.3 Removal of Contents .............................................................. 7-6 7.3 Preparation of Empty Package for Transport ....................................... 7-8 7.4 References ......................................................................................... 7-10 Acceptance Tests................................................................................. 8-1 Visual Inspection..................................................................... 8-1 Structural and Pressure Tests ................................................ 8-2 Containment Boundary Leakage Tests ................................... 8-3 Component Tests ................................................................... 8-4 Shielding Tests ....................................................................... 8-5 Neutron Absorber Tests .......................................................... 8-6 Thermal Acceptance Tests ..................................................... 8-6 Maintenance Program .......................................................................... 8-6 Structural and Pressure Tests ................................................ 8-6 Leak Tests .............................................................................. 8-7 Subsystem Maintenance ........................................................ 8-7 Shielding ................................................................................. 8-8 Thermal .................................................................................. 8-8 References ........................................................................................... 8-9 Page TOC-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 1 GENERAL INFORMATION TABLE OF CONTENTS Introduction ............................................................................................. 1-1 Package Description ............................................................................... 1-1 Packaging .................................................................................... 1-1 Operational Features ................................................................... 1-7 Contents of Packaging ................................................................. 1-8 References ........................................................................................... 1-10 Appendices ........................................................................................... 1-11 TN-40 Packaging Drawings ....................................................... 1-11 LIST OF TABLES Table 1-1 Nominal Dimensions and Weights of the TN-40 Packaging.................. 1-12 Table 1-2 Fuel Qualification Table ........................................................................ 1-13 LIST OF FIGURES Figure 1-1 General Arrangement ........................................................................... 1-14 Figure 1-2 TN-40 Loading Curve (with and without BPRAs) .................................. 1-16 Page 1-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 1.0 GENERAL INFORMATION Introduction This Safety Analysis Report (SAR) presents the evaluation of a Type B(U) spent fuel transport packaging developed by Transnuclear, Inc. and designated the TN-40. This SAR describes the design features and presents the safety analyses which demonstrate that the TN-40 complies with applicable requirements of 10 CFR 71 [1]. The format and content of this SAR follow the guidelines of Regulatory Guide 7.9 [2].

The TN-40 is a dual purpose cask intended for both storage and transport. The TN-40 is currently licensed for storage at the Prairie Island Nuclear Generating Plant (Docket No. 72-0010). A separate storage SAR was submitted by Prairie Island in support of the storage license application. It addresses the safety related aspects of storing spent fuel in TN-40 casks in accordance with 10 CFR 72 [3].

The TN-40 cask is to be licensed for a one-time use. That is, for shipment of the spent fuel it contained during storage. This one-time use is defined to include any sequence of shipments as long as the spent fuel originally loaded has not been removed. The packaging is intended to be shipped as exclusive use. The Criticality Safety Index (CSI) for nuclear criticality control for the TN-40 cask is determined to be zero (0) in accordance with 10 CFR 71.59 [1]. See Chapter 6 for details of this determination.

Transnuclear, Inc. has an NRC approved quality assurance program (Docket Number 71-0250) which satisfies the requirements of 10 CFR 71 Subpart H [1].

Package Description Packaging The TN-40 packaging will be used to transport up to 40 PWR undamaged fuel assemblies with or without fuel inserts. In its transport configuration, the TN-40 packaging consists of the following components:

A basket assembly which locates and supports the fuel assemblies, transfers heat to the cask inner shell, and provides sufficient neutron absorption to satisfy nuclear criticality requirements.

A containment vessel including a closure lid and metallic seals which provides radioactive materials containment and maintains an inert gas atmosphere.

A thick-walled, forged steel gamma shield shell, bottom shield and lid shield plate provide shielding that surrounds the containment vessel.

A radial neutron shield surrounding the gamma shield shell which provides additional radiation shielding. The neutron shielding is enclosed in a steel outer shell.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

A set of impact limiters consisting of balsa and redwood, encased in stainless steel shells, which are attached to either end of the cask body during the shipment. An aluminum spacer is also present to provide a smooth contact surface between the top impact limiter and the cask lid. The impact limiters are held in place with tie rods and attachment bolts.

Sets of upper and lower trunnions that provide support, lifting, and rotation capability for the cask.

A personnel barrier is mounted to the transport frame to prevent unauthorized access to the cask body. The overall dimensions of the TN-40 packaging are 260.87 in. long and 144 in. in diameter with the impact limiters installed. The cask body is 183.75 in. long (with the lid installed) and 91 in. in diameter. The lid is 82.75 in. in diameter. The cask outside diameter including the radial neutron shield is 101.0 in. The cask cavity is 163 in. long and 72.0 in. in diameter. The general arrangement of the TN-40 packaging is depicted in Figure 1-1. Detailed design drawings for the TN-40 packaging are provided in Appendix 1.4. The materials used to fabricate the cask are shown in the Parts List on drawing 10421-71-1 and the materials used to fabricate the impact limiters are shown in the Parts List on drawing 10421-71-41. Where more than one material has been specified for a component, the most limiting properties are used in the analyses in the subsequent chapters of this SAR.

The gross weight of the loaded package is 271.5 kips including a payload of 52.0 kips.

Table 1-1 summarizes the dimensions and weights of the TN-40 packaging. Trunnions attached to the cask body are provided for lifting and handling operations, including rotation of the packaging between the horizontal and vertical orientations. The TN-40 packaging is loaded in the vertical configuration and transported in the horizontal orientation on a specially designed shipping frame.

The maximum normal operating pressure of the TN-40 is 15.7 psig as evaluated in Chapter 4. A cask cavity pressure of 100 psig is conservatively used for the purposes of structural analyses. The spent fuel payload is shipped dry in a helium atmosphere.

The heat generated by the spent fuel assemblies is rejected to the surrounding air by convection and radiation. No forced cooling or cooling fins are required.

The following sections provide a physical and functional description of each major component. Engineering drawings showing dimensions of significance to the safety analyses, welding and NDE information, and a complete materials list are provided in Appendix 1.4. Reference to these drawings is made in the following physical description sections and in general, throughout this SAR. Fabrication of the TN-40 packaging is performed in accordance with these drawings.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 1.2.1.1 Containment Vessel The containment boundary components consist of the inner shell and bottom inner plate, shell flange, lid outer plate, lid bolts, penetration cover plates and bolts (vent and drain) and the inner metallic seals of the lid seal and the vent and drain port cover seals (Figure 4-1). The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. Helium assists in removal of decay heat and provides a non-reactive environment to protect fuel assemblies against fuel cladding degradation which might otherwise lead to gross cladding rupture.

The overall containment vessel length is approximately 170.5 in. with a wall thickness of 1.5 in. The cylindrical cask cavity has a nominal diameter of 72.0 in. and a length of 163 in. The lid outer plate is 4.5 in. thick and is fastened to the body by 48 lid closure bolts. Double metallic seals are provided for the lid closure. To preclude air in-leakage, the cask cavity is pressurized with helium above atmospheric pressure.

The cask cavity can be accessed using two penetrations through the lid. These penetrations are for draining and venting. Double metallic seals are utilized to seal these two lid penetrations.

The over-pressure (OP) port provides access to the volumes between the double seals in the lid and cover plates for leak testing purposes. The OP port cover is not part of the containment boundary.

The inner shell and bottom inner plate materials are SA-203, Grade E or Grade D. The lid outer plate material is SA-350, Grade LF3 or SA-203 Grade E. The TN-40 containment vessel is designed, fabricated, examined and tested in accordance with the requirements of Subsection NB [4] of the ASME Code to the maximum practical extent.

In addition, the design meets the requirements and Regulatory Guides 7.6 [5] and 7.8

[6]. Alternatives to the ASME Code are discussed in Section 2.11. The construction of the containment boundary is shown on drawings 10421-71-3, 4 and 5 provided in Appendix 1.4. The design of the containment boundary is discussed in Chapter 2 and the fabrication requirements (including examination and testing) of the containment boundary are discussed in Chapter 4 and Chapter 8.

1.2.1.2 Gamma and Radial Neutron Shielding A gamma shield is provided around the inner shell and the bottom inner plate of the containment vessel, by an independent shell and bottom plate of carbon steel (Drawing 10421-71-3). The gamma shield shell completely surrounds the containment vessel inner shell and bottom inner plate. The 8.0 in. thick gamma shield shell and the 8.75 in.

thick bottom shell are SA-105, SA-516 Grade 70, or SA-266 Class 4 material.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 In order to obtain a close fit between the inner shell and the gamma shield shell for heat transfer, the gamma shield shell is heated prior to assembly with the inner shell. As the gamma shield shell cools, an axial gap may form between the shell flange and the top of the gamma shield shell. This gap is filled with shims. The shims are machined to fill the gap and act as a backing plate for the weld between the shell flange and the gamma shield shell.

A 6.0 in. thick shield plate (SA-105 or SA-516, Grade 70) is also welded to the inside of the lid outer plate (drawing 10421-71-4).

Radial neutron shielding is provided by a borated polyester resin compound surrounding the gamma shield shell. The resin compound is cast into long, slender aluminum alloy containers. The total radial thickness of the resin and aluminum is 4.50 in. The array of resin-filled containers is enclosed within a 0.50 in. thick outer steel shell (SA-516, Grade 55 or equivalent) constructed of two half cylinders. In addition to serving as resin containers, the aluminum containers provide a conduction path for heat transfer from the gamma shield shell to the outer shell. A pressure relief valve is mounted on top of the resin enclosure to limit the internal pressure increase that may be caused by heating of the resin enclosure for hypothetical accident conditions.

The resin material is an unsaturated polyester cross-linked with styrene, with approximately 50 weight % mineral and fiberglass reinforcement. The components are polyester resin, styrene monomer, alpha methyl styrene, aluminum oxide, zinc borate, and chopped fiberglass which produce the nominal elemental resin composition shown below.

Element wt%

H 5.05 B 1.05 C 35.13 Al 14.93 O + Zn (balance) 43.84 The resin used for the radial neutron shield is a proprietary formulation that has been utilized for the TN-40, TN-32, and TN-68 casks which have been licensed for storage.

The TN-68 cask is also licensed for transport. Information on the resin has been provided to the NRC in support of those licensing applications. The average measured hydrogen weight percent in the resin for TN-40 casks is 5.21. Appendix 9A of the TN-68 storage FSAR provides information on the resin. Information from Appendix 9A of the TN-68 storage FSAR is provided below.

The resin material is an unsaturated polyester crosslinked with styrene, with about 50 weight % mineral and fiberglass reinforcement. The components are polyester resin, styrene monomer, alpha methyl styrene, aluminum oxide, zinc borate, and chopped fiberglass.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Thermal Stability Thermal aging tests on a material with the same components in slightly different proportions have been performed by Transnucleaire, Paris (TNP). The tests by TNP evaluate weight loss and off-gassing at 260 °F (125 °C) and 311 °F (155 °C). An exponential weight loss occurs that rapidly approaches a maximum value. After 106 hours0.00123 days 0.0294 hours 1.752645e-4 weeks 4.0333e-5 months , the weight loss is about 1.0%, and extrapolation of the results indicates maximum weight loss of about 1.3%. This effect diminishes rapidly with decreasing temperature. An analysis of the gas released from a sample heated from 25 to 125 °C over one hour shows it to be 99.9% styrene.

These results obtained with small samples (50 mm thick x 50 mm dia) are conservative with respect to the material in a larger enclosed form such as the TN-40 radial neutron shield, where volatile constituents must diffuse through a much greater distance to be released.

Radiation Stability The European Organization for Nuclear Research (CERN) has published a compilation of its own testing and of prior published data on the radiation resistance of various materials. The data show that while unfilled polyester has poor radiation resistance, both mineral- and glass-filled polyester, such as used in the TN-40 radial neutron shield, are among the most radiation-resistant of thermosetting resins.

In addition, the maximum normal temperature in the TN-40 radial neutron shield is 233 °F (111 °C) at the beginning of storage per Chapter 3 of the TN-40 SAR and for NCT the resin temperature is 229 °F (109 °C). These temperatures are bounded by the temperatures in the thermal aging tests. Therefore, the resin degradation and weight loss are expected to be bounded by the test results.

There are over 100 TN-40, TN-32, and TN-68 casks currently in storage at various ISFSI sites in the United States. Periodic inspections and dose rate measurements of the casks in storage have not indicated any evidence of deterioration of the neutron shielding.

Furthermore, prior to transport of the TN-40 packaging, dose rate measurements must be performed to demonstrate that they meet the 10CFR71.47 criteria. This assures that the resin material has retained adequate properties to meet transport requirements.

The structural analysis of the TN-40 cask shielding is presented in Chapter 2.

Noncontainment welds are inspected in accordance with the NDE acceptance criteria of ASME B&PV Code Subsection NF [7].

Page 1-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 1.2.1.3 Impact Limiters Top (front) and bottom (rear) impact limiters, shown on drawings 10421-71-2 and -40 through -44, form a part of the TN-40 packaging. The impact limiters are attached to each other using 13 tie rods and to the cask by bolt attachment brackets welded to the outer shell in eight locations (four bolting locations per impact limiter). The impact limiters consist of balsa wood and redwood blocks, encased in sealed stainless steel shells (A-240, Grade 304) that maintain a dry atmosphere for the wood and confine the wood when crushed during a free drop. The impact limiters have internal radial gussets for added strength and confinement.

The impact limiters have an outside diameter of 144 in., and an inside diameter of 92 in.

to accommodate the cask ends. The bottom limiter is notched to fit over the lower trunnions. The impact limiters extend axially 37.75 in. from either end of the cask, and overlap the sides of the cask by 12.25 in.

Thirteen 1.5 in. diameter tie-rods are used to hold the impact limiters in place. The tie-rods span the length of the cask and connect to both impact limiters via mounting brackets (See drawings 10421-71-40, and 10421-71-44). The impact limiters are also attached to the outer shell of the cask with eight 1.5 in. diameter bolts. The bolts are inserted through brackets (welded to the cask outer shell) and thread into each impact limiter. There are a total of eight bracket sets, four per impact limiter.

Each impact limiter is provided with nine fusible plugs that are designed to melt during a fire accident, thereby relieving excessive internal pressure. Each impact limiter has two lifting lugs for handling, and two support angles for holding the impact limiter in a vertical position during storage. The lifting lugs and the support angles are welded to the stainless steel shells.

An aluminum spacer is placed on the cask lid prior to mounting the top impact limiter.

The purpose of the aluminum spacer is to provide a smooth contact surface between the lid and the top impact limiter. The top plate of the spacer has 48 holes to allow clearance for the lid bolt heads. The lip of the spacer is designed to make up the difference between the lid and cask outer diameters so that the top impact limiter cavity mates with a surface of constant diameter (drawing 10421-71-7).

The functional description as well as the performance analysis of the impact limiters is provided in Appendix 2.10.8.

1.2.1.4 Tiedown and Lifting Devices Threaded holes are provided in the lid for attachment of component lifting devices.

These are used as attachment points for sling systems or other lifting tools. These threaded holes are equally spaced 90° apart as shown on drawing 10421-71-4. Prior to transport, any attachments will be removed. Access to these threaded holes is prevented by the presence of the top impact limiter.

Page 1-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Four trunnions, which form part of the cask body, are attached for lifting and rotating of the cask. Two of the trunnions are located near the top of the body, and two near the bottom.

The upper trunnions are welded to the gamma shield shell and are designed to meet the requirements of 10CFR 71.45(a). This is accomplished by evaluating the trunnions to the stress design factors of 6 and 10 when compared with yield and ultimate stress limits respectively. These design loads exceed the 10CFR 71.45(a) requirements. The lower trunnions are welded to the gamma shield shell and bottom shield, and are used for rotating the cask between the vertical and the horizontal positions.

The tiedown devices are described in Section 2.5.2 of Chapter 2.

1.2.1.5 Fuel Basket The basket structure is designed, fabricated and inspected in accordance with ASME B&PV Code Subsection NB [4]. Section 2.1.2.2 of Chapter 2 discusses use of NB instead of NG. Alternatives to the NB code are provided in Section 2.11. The basket structure consists of an assembly of stainless steel cells joined by a fusion welding process and separated by aluminum and poison plates which form a sandwich panel.

The panel consists of two aluminum plates which sandwich a poison plate. The aluminum plates provide the heat conduction paths from the fuel assemblies to the cask inner plate. The poison material provides the necessary criticality control. This method of construction forms a very strong honeycomb-like structure of cell liners which provide compartments for 40 fuel assemblies. The open dimension of each cell is 8.05 in. x 8.05 in. which provides a minimum of 1/8 in. clearance around the fuel assemblies. The overall basket length (160.0 in.) is less than the cask cavity length to allow for thermal expansion and fuel assembly handling.

Operational Features There are no complex operational features associated with the TN-40 packaging. The TN-40 cask and basket are designed to be compatible with spent fuel pool loading/unloading methods. The sequential steps to be followed for cask loading, testing, and unloading operations are provided in Chapter 7. Chapter 7 also provides criteria and limits for operational tests. The loading operations are summarized below.

Upon arrival, the empty cask is inspected. Preparation of the packaging for loading/unloading requires that the top (front) and bottom (rear) impact limiters including the tie-rods and attachment bolts are first removed from the cask. The cask is lifted from the transport frame to an upending/downending frame. The cask is then rotated from the horizontal transport orientation to the vertical orientation using a crane and lift beam attached to the upper trunnions. The lower trunnions pivot in the upending/downending frame as the cask is rotated.

Page 1-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The cask is brought into the spent fuel building. Access to the cask cavity and fuel basket is obtained by untorquing and removing the 48 closure lid bolts, and removing the lid using hoist rings threaded into the lid. The cask is then lowered into the cask pit/spent fuel pool. Fuel assemblies are loaded into the 40 basket compartments.

The lid is installed and the cavity is vented. The cask is lifted so that the lid is above the surface of water and some of the lid bolts are installed hand tight. The cask may be drained at this time, or after removal from the pool. The cask is moved from the cask pit/spent fuel pool to the decontamination area. The remaining lid bolts are installed and tightened to the specified torque. The cask cavity is then dried by means of a vacuum system and then back-filled with helium. The lid seals and penetration cover seals are leak tested. The external surface radiation levels are checked to assure that they are within limits.

Contents of Packaging The characteristics of the contents of the TN-40 packaging are limited to the following:

a. Fuel shall be unconsolidated;

b. Fuel shall only have been irradiated at the Prairie Island Nuclear Generating Plant Unit 1 cycles 1 through 16 or Unit 2 cycles 1 through 15;

c. Fuel shall be limited to fuel types:

i. Westinghouse 14X14 Standard, ii. Exxon 14X14 Standard (includes high burnup standard),

iii. Exxon 14X14 TOPROD, and iv. Westinghouse 14X14 OFA;

d. Fuel may include burnable poison rod assemblies (BPRAs) provided:

i. the BPRA has cooled for 25 years, ii. the average cumulative exposure of the fuel assembly(s) where the BPRA(s) resided during reactor operation shall be 30,000 MWd/MTU;

e. Fuel may include thimble plug assemblies (TPAs) provided:

i. the TPA has cooled for a minimum of 25 years, ii. the average cumulative exposure of the fuel assembly(s) where the TPA(s) resided during reactor operation shall be 125,000 MWd/MTU;

f. The combined weight of a fuel assembly and any BPRA or TPA shall be

< 1,330 lbs;

g. The combined weight of all fuel assemblies, BPRAs, and TPAs in a single cask shall be < 52,000 lbs;

h. The number of assemblies in the container shall be < 40;

i. The initial enrichment shall be 3.85 weight percent U-235;

j. The assembly average burnup shall be 45,000 MWd/MTU;

k. The assembly average burnup shall be the loading curve shown in Figure 1-2; Page 1-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

l. The minimum cooling time for various combinations of minimum assembly average enrichment and maximum assembly average burnup prior to transport shall be in accordance with Table 1-2, and

m. The fuel assemblies shall not be Unit 1 Region 4 fuel assemblies (i.e.,

assemblies identified as D-01 through D-40).

n. The maximum uranium loading per fuel assembly is 0.410 MTU.

o. The fuel shall not be a DAMAGED FUEL ASSEMBLY.

A DAMAGED FUEL ASSEMBLY is a spent nuclear fuel assembly that:

is a partial fuel assembly, that is, a fuel assembly from which fuel pins are missing unless dummy fuel pins are used to displace an amount of water equal to that displaced by the original pins; or

has known or is suspected to have structural defects or gross cladding failures (other than pinhole leaks) sufficiently severe to adversely affect fuel handling and transfer capability.

p. The characteristics of the specific fuel types authorized for shipment in the TN-40 Cask are provided in the table below. The table shows the pre-irradiated nominal design dimensions and specifications for the fuel.

Exxon Exxon Exxon Westinghouse Westinghouse Standard Toprod High Burnup Standard OFA Fuel Designations (14x14) (14x14) (14 x 14) (14x14) (14x14)

Rod Pitch (in.) 0.556 0.556 0.556 0.556 0.556 Pellet OD (in.) 0.3565 0.3505 0.3565 0.3659 0.3444 Clad OD (in.) 0.424 0.426 0.417 0.422 0.400 Clad Thickness (in.) 0.0300 0.0295 0.0310 0.0243 0.0243 Number of Fueled Rods 179 179 179 179 179 Clad Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Number of Guide Tubes 16 16 16 16 16 Number of Instrument Tubes 1 1 1 1 1 Active Fuel Length (in.) 144 144 144 144 144 Maximum Length (Assembly+BPRA) (in.) 161.3 161.3 161.3 161.3 161.3 Maximum Width (in.) 7.763 7.763 7.763 7.763 7.763

q. The maximum heat load is 19.0 kW per cask and 0.475 kW per fuel assembly, including the BPRAs, and TPAs.

Page 1-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. 10 CFR 71, Packaging and Transportation of Radioactive Material.

2. USNRC Regulatory Guide 7.9, Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material, Rev. 2, March 2005.

3. 10 CFR 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.

4. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, 1989 without addenda.

5. USNRC Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessel, Rev. 1, March 1978.

6. USNRC Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Cask, Rev. 1, March 1989.

7. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Subsection NF, 1989 without addenda.

Page 1-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Appendices TN-40 Packaging Drawings The following Transnuclear drawings are enclosed:

Drawing No Title 10421-71-1 Rev. 6 TN-40 Transport Packaging Parts List and Notes (1 sheet) 10421-71-2 Rev. 3 TN-40 Transport Packaging Transport Configuration (2 sheets) 10421-71-3 Rev. 3 TN-40 Transport Packaging General Arrangement (1 sheet) 10421-71-4 Rev. 0 TN-40 Transport Packaging Lid Assembly and Details (1 sheet) 10421-71-5 Rev. 0 TN-40 Transport Packaging Lid Details (1 sheet) 10421-71-6 Rev. 0 TN-40 Transport Packaging Trunnion, Basket Rail and Neutron Shield Details (1 sheet) 10421-71-7 Rev. 3 TN-40 and TN-40HT Transport Packaging Impact Limiter Spacer Details (1 sheet) 10421-71-8 Rev. 0 TN-40 Transport Packaging Basket Assembly (1 sheet) 10421-71-9 Rev. 0 TN-40 Transport Packaging Basket Details (1 sheet) 10421-71-40 Rev. 2 TN-40 Transport Packaging Impact Limiters General Arrangement (1 sheet) 10421-71-41 Rev. 2 TN-40 and TN-40HT Transport Packaging Impact Limiters Parts List and Notes (1 sheet) 10421-71-42 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Assembly (1 sheet) 10421-71-43 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Details (1 sheet) 10421-71-44 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Parts (1 sheet)

Page 1-11

Proprietary and Security Related Information for Drawing 10421-71-01, Rev. 6 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-02, Rev. 3 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-03, Rev. 3 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-04, Rev. 0 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-05, Rev. 0 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-06, Rev. 0 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-07, Rev. 3 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-08, Rev. 0 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-09, Rev. 0 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-40, Rev. 2 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-41, Rev. 2 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-42, Rev. 1 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-43, Rev. 1 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 10421-71-44, Rev. 1 Withheld Pursuant to 10 CFR 2.390

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 1-1 Nominal Dimensions and Weights of the TN-40 Packaging Overall length (with impact limiters, in.) 261 Overall length (without impact limiters, in.) 184 Impact Limiter Outside diameter (in.) 144 Outside diameter (without impact limiters, in.) 101 Cavity diameter (in.) 72.0 Cavity length (in.) 163 Containment shell thickness (in.) 1.5 Containment vessel length (in.) 170.5 Body wall thickness (in.) 9.5 Containment Lid thickness (in.) 4.5 Overall Lid thickness (in.) 10.5 Bottom thickness (in.) 10.3 Resin and aluminum box thickness (in.) 4.5 Outer shell thickness (in.) 0.5 Overall basket length (in.) 160 Weight of Fuel Assemblies (with inserts) (kips) 52.0 Loaded Weight of TN-40 Cask (without impact limiters) (kips) 236.5 Weight of Impact Limiters, Aluminum Spacer, and Tie-Rods (kips) 35.0 Total Loaded Weight of TN-40 Packaging (w/o shipping frame) (kips) 271.5 Page 1-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 1-2 Fuel Qualification Table MINIMUM COOLING TIMES (YEARS)

Minimum Assembly Average Enrichment (wt. % U235)

Maximum Assembly Average Burnup (GWD/MTU) 2 2.25 2.35 2.75 3 3.25 3.4 3.6 3.85 17 30 30 30 30 30 30 30 30 30 18 30 30 30 30 30 30 30 30 30 19 30 30 30 30 30 30 30 30 30 20 30 30 30 30 30 30 30 30 30 21 30 30 30 30 30 30 30 30 30 22 30 30 30 30 30 30 30 30 30 23 30 30 30 30 30 30 30 30 30 24 30 30 30 30 30 30 30 30 30 25 30 30 30 30 30 30 30 30 30 26 30 30 30 30 30 30 30 30 30 27 30 30 30 30 30 30 30 30 30 28 30 30 30 30 30 30 30 30 30 29 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 31 30 30 30 30 30 30 30 32 30 30 30 30 30 30 30 33 30 30 30 30 30 30 30 34 30 30 30 30 30 30 30 35 30 30 30 30 30 30 30 36 30 30 30 30 30 30 30 37 30 30 30 30 30 30 30 38 30 30 30 30 30 30 30 39 30 30 30 30 30 30 30 40 30 30 30 30 30 30 30 41 30 30 30 30 30 30 30 42 30 30 30 30 30 30 30 43 30 30 30 30 30 44 30 30 30 30 45 30 30 30 30 Note:

1. For fuel characteristics that fall between the assembly average enrichment values in the table, use the next lower enrichment, and next higher burnup to determine minimal cooling time.

2. Fuel assemblies that were located in the RCCA control bank D position during Unit 1 cycle 1 and Unit 2 cycle 1 shall have a minimum cooling time of greater than 35 years.

Page 1-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 1-1 General Arrangement TN-40 PACKAGING Page 1-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 1-1 General Arrangement TN-40 PACKAGING Notes to Figure 1-1 A. Some details exaggerated for clarity.

B. Components are listed below:

1 Top (front) Impact Limiter 2 Bottom (rear) Impact Limiter 3 Top Impact Limiter Spacer 4 Tie Rod 5 Impact Limiter Bolting & Bracket 6 Upper Trunnions 7 Lower Trunnions 8 Cask Body (Gamma Shield Shell & Bottom Shield) 9 Containment Shell (Inner Shell & Bottom Inner Plate) 10 Radial Neutron Shielding 11 Outer Shell 12 Lid Assembly 13 Drain Tube Page 1-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 POLYNOMIAL FITS FOR LOADING CURVES 38000 36000 34000 32000 30000 y = -1259.8x2 + 20242x - 23617 R2 = 0.9956 28000 BURNUP (MWD/MTU) 26000 24000 22000 20000 y = - 366.95x2 + 14770x - 17200 R2 = 0.9966 18000 BPRA-CURVE 16000 NOBP-CURVE 14000 12000 10000 2.00 2.20 2.40 2.60 2.80 3.00 3.20 3.40 3.60 3.80 4.00 ENRICHMENT (WT. % U-235)

Note:

1.) The ACCEPTABLE region for the BPRA-CURVE always lies above the curve and corresponds to a burnup that is greater than or equal to the burnup loading curve. The ACCEPTABLE region for the NOBP-CURVE always lies above the curve and corresponds to a burnup that is greater than or equal to the burnup loading curve.

2.) The maximum initial enrichment within the fuel assembly, e.g., the central region of the fuel assembly, shall be used when applying this curve.

Figure 1-2 TN-40 Loading Curve (with and without BPRAs)

Page 1-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 2 STRUCTURAL EVALUATION TABLE OF CONTENTS Structural Design .................................................................................... 2-1 Discussion .................................................................................... 2-1 Design Criteria ............................................................................. 2-3 Weights and Center-Of-Gravity ............................................................... 2-6 Mechanical Properties of Materials ......................................................... 2-6 Cask Material Properties .............................................................. 2-6 Basket Material Properties ........................................................... 2-7 Impact Limiter Material Properties................................................ 2-7 Fracture Toughness Requirements .............................................. 2-7 General Standards For All Packages ...................................................... 2-7 Minimum Package Size ................................................................ 2-7 Tamper-proof Feature .................................................................. 2-7 Positive Closure ........................................................................... 2-7 Chemical and Galvanic Reactions ............................................... 2-8 Lifting And Tie-Down Standards ........................................................... 2-10 Lifting Devices ............................................................................ 2-10 Tie-Down Devices ...................................................................... 2-16 Normal Conditions Of Transport ........................................................... 2-17 Heat ........................................................................................... 2-18 Cold Environment....................................................................... 2-19 Increased External Pressure (N4) .............................................. 2-20 Reduced External Pressure (N5) ............................................... 2-20 Transport Shock Loading (N14 & N15) ...................................... 2-20 Transport Vibration Loading (N12 & N13) .................................. 2-21 Water Spray ............................................................................... 2-21 Free Drop (N6 through N11) ...................................................... 2-21 Corner Drop ............................................................................... 2-22 Compression .............................................................................. 2-22 Penetration ................................................................................. 2-22 Lid Bolt Analysis ......................................................................... 2-22 Fatigue Analysis of the Containment Boundary ......................... 2-22 Page 2-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Structural Evaluation of the Basket under Normal Condition Loads ......................................................................... 2-26 Summary of NCT Cask Body Structural Analysis....................... 2-26 Hypothetical Accident Conditions ......................................................... 2-27 30 Foot Free Drop ...................................................................... 2-28 Puncture ..................................................................................... 2-33 Thermal ...................................................................................... 2-36 Water Immersion ........................................................................ 2-37 Structural Evaluation of the Basket under Accident Loads ......... 2-39 Summary of HAC Cask Body Structural Analysis ...................... 2-40 Special Forms / Fuel Rods.................................................................... 2-40 Special Form .............................................................................. 2-40 Fuel Rods ................................................................................... 2-40 References ........................................................................................... 2-41 Appendices ........................................................................................... 2-43 ASME Code Alternatives ...................................................................... 2-44 Page 2-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 2-1 Evaluation Method Employed to Demonstrate Compliance With Specific Regulatory Requirements ........................................................ 2-49 Table 2-2 Containment Vessel Stress Limits ........................................................ 2-50 Table 2-3 Cover Bolt Stress Limits........................................................................ 2-51 Table 2-4 Non Containment Structure Stress Limits ............................................. 2-52 Table 2-5 Basket Stress Limits ............................................................................. 2-53 Table 2-6 Cask Weight and Center of Gravity ...................................................... 2-54 Table 2-7 Trunnion Section Properties and Applied Loads ................................... 2-55 Table 2-8 Trunnion Stresses when Loaded by 6 and 10 Times Cask Weight (Lifting) ...................................................................................... 2-56 Table 2-9 TN-40 Performance Evaluation Overview (Normal Conditions of Transport) ............................................................................................. 2-57 Table 2-10 Individual Load Cases for Normal Conditions of Transport TN-40 Cask Body Analysis ......................................................................... 2-58 Table 2-11 Summary of Load Combinations for Normal Condition of Transport .............................................................................................. 2-59 Table 2-12 Reference Temperatures for Stress Analysis Acceptance Criteria .................................................................................................. 2-61 Table 2-13 Summary of Load Combination Stresses for Normal Conditions of Transport .......................................................................................... 2-62 Table 2-14 Linearized Stress Evaluation of Normal Condition of Transport Load Combinations ............................................................................... 2-64 Table 2-15 TN-40 Performance Evaluation Overview (Hypothetical Accident Conditions of Transport) ....................................................................... 2-68 Table 2-16 Summary of Individual Load Factors for Hypothetical Accident Condition of Transport .......................................................................... 2-69 Table 2-17 Summary of Load Combinations for Hypothetical Accident Condition of Transport .......................................................................... 2-70 Table 2-18 Summary of Load Combination Stresses for Hypothetical Accident Condition of Transport ............................................................ 2-71 Table 2-19 Linearized Stress Evaluation for Hypothetical Accident Condition Load Combinations ............................................................... 2-72 Page 2-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 2-1 Geometry of Upper (front) and Lower (rear) Trunnions......................... 2-74 Figure 2-2 290 psig Immersion Analysis Finite Element Model Loads and Boundary Conditions ............................................................................ 2-75 Figure 2-3 DELETED ................................................................................................. 2-76 Figure 2-4 DELETED ................................................................................................. 2-76 Figure 2-5 DELETED ................................................................................................. 2-76 Figure 2-6 DELETED ................................................................................................. 2-76 Page 2-iv

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.0 STRUCTURAL EVALUATION Structural Design This chapter, including its appendices, presents the structural evaluation of the TN-40 transport packaging. This evaluation consists of numerical analyses and impact limiter testing which demonstrate that the TN-40 packaging satisfies applicable requirements for a Type B(U) packaging.

Discussion The structural integrity of the packaging under normal conditions of transport (NCT) and hypothetical accident conditions (HAC) specified in 10CFR71 [1] is shown to meet the design criteria described in Section 2.1.2. The TN-40 transport package consists of three major structural components: the cask body, the fuel basket, and the impact limiters (top and bottom). These components are described in Chapter 1 and are shown on drawings provided in Appendix 1.4.1.

The cask body is described in Section 1.2. Drawing 10421-71-1 shows the parts list.

Drawing 10421-71-2 shows the overall transport configuration of the TN-40 transport package. Drawing 10421-71-3 shows the general arrangement of the TN-40 packaging.

Drawings 10421-71-4 and 10421-71-5 present the lid assembly and details. Drawing 10421-71-6 shows the trunnion/basket rail/neutron shield details and 10421-71-7 shows the impact limiter top spacer. Drawings 10421-71-8 and 10421-71-9 show the basket assembly. Drawings 10421-71-40 through 10421-71-44 provide details of the impact limiter design.

The inner shell and the bottom inner plate are made of SA-203, Grade D or E. The shell flange is SA-350 Grade LF3 and the lid outer plate is constructed with SA-350 Grade LF3 or SA-203 Grade E. The gamma shield shell and bottom shield are SA-266, CL 4, SA-516, Grade 70, or SA-105. The lid shield plate is constructed from SA-105 or SA-516 Grade 70.

In order to obtain a close fit between the inner shell and the gamma shield shell, for better heat transfer, the gamma shield shell is heated prior to assembling it with the inner shell. As the gamma shield shell cools, an axial gap may form between the shell flange and the gamma shield. Shims may be machined to fill the gap and act as a backing plate for the 0.50 inch weld, between the shell flange and the gamma shield shell.

The four upper and lower trunnions are cylindrical, SA-105 or SA-266 Class 4 forgings that are welded to the gamma shield shell of the cask body. The two upper trunnions are designed to lift the loaded TN-40 cask vertically. The lower trunnions provide the capability to rotate the cask on the upending/downending frame. The trunnions are shown in Drawing 10421-71-6.

Page 2-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The outer shell around the neutron shield consists of a cylindrical shell section, with closure plates at each end of the neutron shield. The closure plates are welded to the outer surface of the gamma shield shell. The outer shell provides an enclosure for the resin-filled aluminum containers, and maintains the resin in the proper location with respect to the active length of the fuel assemblies in the cask cavity. The shell is made of SA516 Grade 55 or equivalent carbon steel.

The basket is a welded assembly of stainless steel fuel compartment boxes, and is designed to accommodate 40 fuel assemblies. The fuel compartment stainless steel box sections are attached together locally by cylindrical stainless steel plugs (that pass through the aluminum and Boral plates) that are fusion welded to the adjacent box sections. The basket contains 40 compartments for proper spacing and support of the fuel assemblies. Neutron poison plates, constructed from Boral, are sandwiched between the sections of the stainless steel walls of the adjacent box and the adjacent stainless steel plates. Drawings 10421-71-8 and 10421-71-9 show details of the basket.

Structural rails oriented parallel to the axis of the cask are attached to the inner surface of the inner shell to establish and maintain basket orientation, to prevent twisting of the basket assembly, and to support the edges of those plates adjacent to the rails, which would otherwise be free to slide tangentially around the cask cavity wall under lateral inertial loadings.

The cask body and the fuel basket together with the two impact limiters, form the packaging which is designed to meet all of the applicable 10CFR71 [1] requirements for a Type B(U) packaging. The maximum normal operating pressure (MNOP) is 15.7 psig.

The wall thickness of the cask body (excluding the outer shell and outer shell closure plates) enables the packaging to withstand the hypothetical puncture accident. The gamma shield shell is designed to be both strong and ductile. The top and bottom impact limiters absorb the kinetic energy for the 1 foot NCT and 30 foot HAC free drops.

Table 2-1 summarizes the specific evaluation methods that are used to demonstrate compliance with the regulations. Numerical analyses have been performed for the NCT and HAC event, as well as for the lifting and tie-down loads. In general, numerical analyses have been performed for all of the regulatory events. These analyses are summarized in the main body of this section, and are described in detail in Appendices 2.10.1 through 2.10.8. Testing of one-third scale impact limiters has been performed to provide the design basis impact loads used in the TN-40 structural analyses. The results of the testing are provided in Appendix 2.10.9.

The detailed structural analyses of the TN-40 packaging are included in the following appendices:

Appendix 2.10.1 Structural Analysis of the Cask Body

Appendix 2.10.2 Lid Bolt Analysis Page 2-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

Appendix 2.10.3 Structural Analysis of the Outer Shell

Appendix 2.10.4 Fracture Toughness Evaluation of the TN-40 Cask

Appendix 2.10.5 Structural Analysis of the TN-40 Basket

Appendix 2.10.6 Dynamic Load Factor for Basket Drop Analysis

Appendix 2.10.7 Structural Evaluation of the Fuel Rod Cladding Under Accident Impact

Appendix 2.10.8 Structural Evaluation of the Impact Limiters

Appendix 2.10.9 TN-40 Impact Limiter Testing

Appendix 2.10.10 Inner Shell Buckling Due to External Pressure

Appendix 2.10.11 TN-40 Lid Closure Evaluation Due to Delayed Impact Design Criteria The packaging consists of three major components:

Cask Body

Fuel Basket

Impact Limiters The structural design criteria for these components are described below.

2.1.2.1 Cask Body 2.1.2.1.1 Containment Vessel The containment vessel consists of the inner shell with the shell flange out to the seal seating surface, the bottom inner plate, and the lid. The lid bolts and seals are also part of the containment vessel as are the drain and vent port cover plates, bolts and seals.

The containment vessel is designed to the maximum practical extent as an ASME Class I component in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB [3]. The Subsection NB rules for materials, design, fabrication and examination are applied to all of the above components to the maximum practical extent. In addition, the design meets the requirements of Regulatory Guides 7.6 [5] and 7.8 [6]. Alternatives to the ASME Code are discussed in Section 2.11 of this Chapter.

The acceptability of the containment vessel under the applied loads is based on the following criteria:

Title 10, Chapter 1, Code of Federal Regulations, Part 71

Regulatory Guide 7.6 Design Criteria

ASME Code Design Stress Intensity Limits Page 2-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

Preclusion of Fatigue Failure

Preclusion of Brittle Fracture The stresses due to each load are categorized as to the type of stress induced (e.g.,

membrane, bending) and the classification of stress (e.g., primary, secondary). Stress limits for containment vessel components, other than bolts, for NCT and HAC loads are given in Table 2-2. The stress limits used for HAC conditions, determined on an elastic basis, are based on the entire structure (containment vessel and gamma shielding material) resisting the accident load. Local yielding is permitted at the point of contact where the load is applied.

The primary membrane stresses and primary membrane plus bending stresses in the cask are limited under NCT to Sm (Sm is the code allowable stress intensity) and 1.5 Sm, respectively.

The HAC impact events are evaluated as short duration, Level D conditions. The stress criteria are taken from Section III, Appendix F of the ASME Code [7]. For elastic quasi-static analysis, the primary membrane stress intensity Pm is limited to 0.7 Su, and membrane plus bending stress intensities (Pm + Pb) are limited to Su.

The allowable stress limits for the cover bolts are listed in Table 2-3. The allowable stress limits for the lid closure bolts are listed separately in Appendix 2.10.2, Tables 2.10.2-3 and 2.10.2-4.

The allowable stress intensity values Sm or Su as defined by the Code, are taken at the maximum component temperature calculated for each service load condition.

2.1.2.1.2 Non-Containment Structure Certain components of the cask body such as the gamma shield shell, the neutron shield outer shell and the trunnions do not provide containment but do have structural functions. These components (referred to as non-containment structures) are required to withstand the environmental loads, and in some cases share the loads with the containment vessel. The stress limits for these non-containment structures are given in Table 2-4. The gamma shield shell and neutron shield outer shell are designed, fabricated and inspected in accordance with the ASME Code Subsection NF [3], to the maximum practical extent. Structural and structural attachment welds are examined by the liquid penetrant or the magnetic particle method, in accordance with Section V, Article 6 of the ASME Code [8]. The magnetic particle and liquid penetrant examination acceptance standards are in accordance with Section III, Subsection NF, Paragraphs NF-5340 and NF-5350 [3].

The welding procedures, welders and weld operators are qualified in accordance with Section IX of the ASME Code [10].

The radial neutron shield, including the carbon steel outer shell, has not been designed to withstand all of the HAC loads.

Page 2-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.1.2.2 Basket The basket is designed in accordance with the ASME Code Subsection NB [3], to the maximum practical extent. The TN-40 was developed as a storage cask and as such the basket and containment were designed using the stress limits from ASME Code Subsection NB. NCT stress limits specified in NB are the same as NG which is currently used for transportation packages. For HAC, both NB and NG require use of Appendix F for the stress limits. Therefore, the basket design meets the NG stress limits as specified in the current Standard Review Plan [12]. The alternatives to ASME codes are listed in Section 2.11.

The neutron poison sheets are not included in the structural analysis. Therefore, the materials are not required to be ASME Code materials. The aluminum plates between the fuel compartments and aluminum basket rail are not Class 1 material. Aluminum was selected for its excellent thermal conductivity and a high strength to weight ratio.

NUREG-3854 [11] and NUREG-1617 [12] allow materials other than ASME Code materials to be used in the cask fabrication. The ASME Code does provide the material properties for the aluminum alloy and also allows the material to be used for Section III applications (Class 2 or 3).

The stress limits for the basket are summarized in Table 2-5. The wall thickness of the basket fuel compartment is designed to meet the heat transfer, nuclear criticality, and the structural requirements. The basket structure provides sufficient rigidity to maintain a subcritical configuration under the applied loads.

The basis for the allowable stresses for the compartment box and the fusion welds isSection III, Division I, Subsection NB of the ASME Code [3]. The primary membrane stresses and primary membrane plus bending stresses in the basket are limited to Sm (Sm is the code allowable stress intensity) and 1.5 Sm, respectively, for NCT loads.

The HAC events are evaluated as short duration, Level D conditions. The stress criteria are taken from Section III, Appendix F of the ASME Code [7]. The membrane and membrane plus bending stresses were compared against 0.7 Su and 0.9 Su elastic-plastic analysis stress criteria values for the HAC drop events.

The fuel compartment walls under compressive loads are also evaluated to ensure that buckling will not occur. ANSYS nonlinear buckling analysis is used to calculate the buckling load. See Appendix 2.10.5 for complete details of criteria for these conditions.

The fusion welds in the basket are not code welds (alternative to codes are listed in Section 2.11). The fusion welds are qualified by testing. The testing program ensures that the fusion welds are stronger than the base metal. Additionally it is shown that the maximum stress in the base metal is lower than the ASME NB (identical to NG) allowable stress thus basket integrity is maintained and the welds are qualified. The testing program is provided in Appendix 2.10.5, Section 2.10.5.4.

Page 2-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.1.2.3 Impact Limiters (Top and Bottom)

The TN-40 packaging includes impact limiters at each end of the cask body. The limiters are nearly identical. The inside diameter of the limiter is determined by the diameter of the gamma shield shell. The length and outside diameter of the limiter are sized to limit the cask inertial loads during the 1 foot NCT and 30 foot HAC drop events so that the containment vessel (and the non-containment structures) meets the design criteria.

The impact limiter stainless steel cylinders, gussets, and end plates are designed to position and confine the balsa and redwood blocks so that the impact energy is properly absorbed. The stainless steel shell is also designed to support and protect the wood blocks under NCT environmental conditions (moisture, pressure, temperature, etc.).

The impact limiters and attachments are designed to withstand the applied loads and to prevent separation of the limiters from the cask during any NCT or HAC impact. The design criteria for the impact limiters and attachments are specified in Appendix 2.10.8.

2.1.2.4 Trunnions The evaluation and design criteria for the lifting trunnions are based on the requirements of 10CFR71.45 [1]. The details of the evaluation are presented in Section 2.5. The evaluation demonstrates that the upper trunnions, used for lifting, have a minimum factor of safety of six against yield or ten against ultimate, whichever is most restrictive. These design loads exceed the 10 CFR 71.45 (a) requirements.

Weights and Center-Of-Gravity The weight of the TN-40 packaging is 271.46 kips. The weights of the major individual subassemblies are listed in Table 2-6. The center of gravity of the cask is located on the axial centerline approximately 91.4 inches from the base of the cask.

The calculations that follow typically use conservative weights that are slightly higher than those listed in Table 2-6. These are:

1. Lifting (w/o impact limiters), 250 kips

2. Tiedown analyses, 271.7 kips

3. Cask Body analysis 271.7 kips Mechanical Properties of Materials Cask Material Properties This section provides the mechanical properties of materials used in the structural evaluation of the TN-40 cask. Drawing 10421-71-1 (see Appendix 1.4.1) lists the materials selected for each component of the transport cask. The minimum yield, ultimate, and design stress values are taken from ASME Code,Section III, Appendices

[7].

Page 2-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Basket Material Properties The material properties of the 304 stainless steel plates are taken from the ASME Code,Section III, Appendices. The material properties of the aluminum alloy (6061-T6) are also taken from the ASME Code [7] and aluminum standards and data [23].

Impact Limiter Material Properties Mechanical properties of the energy absorbing wood and wood adhesive used in the impact limiters are specified in Appendix 2.10.8 (Table 2.10.8-1).

Fracture Toughness Requirements The cask body and closure lid material is a ferritic steel and is therefore subject to fracture toughness requirements in order to assure ductile behavior at the lowest service temperature (LST) of -20° F. See Appendix 2.10.4 for fracture toughness evaluation. The fracture toughness evaluations in Appendix 2.10.4 show that the TN 40 cask materials meet the fracture toughness criteria of NUREG/CR-3826 [21] and NUREG/CR-1815 [22].

General Standards For All Packages The TN-40 is designed to comply with the general standards for all packages specified by 10CFR71.43 [1].

Minimum Package Size The overall package dimensions of 260.87 inches long and 144 inches in diameter exceed the minimum dimension requirement of 10 cm (4 inches).

Tamper-proof Feature The only access path into the package is through the closure lid and associated lid closure bolts. During transport the top (front) impact limiter entirely covers and prevents access to the cask closure lid and the vent and access port penetrations in the lid. A wire security seal is installed in the top (front) impact limiter attachment tierod prior to each shipment. The presence of this seal demonstrates that unauthorized opening of the package has not occurred.

Positive Closure Positive fastening of all access openings through the containment vessel is accomplished by bolted closures which preclude unintentional opening. In addition, the presence of the impact limiters and security seal described in Section 2.4.2 provide further protection against unintentional opening.

Page 2-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Chemical and Galvanic Reactions The materials of the TN-40 cask have been reviewed to determine whether chemical, galvanic or other reactions between the materials, contents and environment might occur during any phase of loading, unloading, handling or transport.

The TN-40 cask components are exposed to the following environments:

During loading and unloading, the casks are submerged in pool water. For PWR plants the pool water is borated. The casks are only kept in the spent fuel pool for a short period of time, typically about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to load or unload fuel. After removing the cask from the pool, water or water vapor is present during the draining and drying process. This takes approximately 1 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to drain, and another 20-24 hours to completely dry, evacuate and backfill the cask with helium.

During handling and transportation, the exterior of the cask is exposed to normal environmental conditions of temperature, rain, snow, etc.

During transportation, the interior of the cask is exposed to an inert helium environment. The helium environment does not support chemical or galvanic reactions because both moisture and oxygen must be present for a reaction to occur. The cask is thoroughly dried by a vacuum drying process before transport. It is then sealed and backfilled with helium.

The radial neutron shielding materials and the aluminum resin boxes are sealed inside the outer shell during normal operations. The resin material is inert after it has cured and does not affect the aluminum boxes or the carbon steel housing.

2.4.4.1 Cask Interior The TN-40 cask materials are shown in the Parts List on Drawing 10421-71-1 (see Appendix 1.4.1).

The containment vessel is made from SA-203 Grade D or E and SA-350 Grade LF3.

The vessel interior surfaces are grit blasted and then metal-sprayed with aluminum/zinc alloy.

The metal-spray coating is subject to the following service environments:

After fabrication, the cask is closed and shipped with air in the cask cavity.

At fuel loading, borated spent fuel pool water is present in the cavity for a short duration.

The cask is vacuum-dried and helium backfilled for storage lifetime of 20 years and/or off-site transport.

At fuel removal, the coating may again be exposed to borated spent fuel pool water for a short duration.

Page 2-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The coating is not subject to abrasion except for the one-time insertion of the basket into the containment vessel.

All sealing surfaces are stainless steel clad by weld overlay. The metallic seals have a stainless steel liner and an aluminum jacket.

Within the cask cavity, there are 6 basket rails made from 6061-T6 or -T651 aluminum.

The rails are shown on the drawings provided in Appendix 1.4. These rails are not coated.

The cask basket is assembled from SA-240, Type 304 stainless steel boxes which are joined together by a fusion welding process and are separated by aluminum and poison plates which form a sandwich panel. The aluminum plates are 6061 -T651 aluminum.

The aluminum plates are held in place by the stainless steel plugs to which the boxes are welded. The aluminum is not welded or bolted to the stainless steel.

The Boral sheets are also held in place by the stainless steel plugs and are captured between the stainless steel boxes. The Boral is not welded or bolted to the stainless steel.

2.4.4.2 Cask Exterior The exterior of the cask is carbon steel. The exterior of the cask, with the exception of the trunnion bearing surfaces is thermal sprayed and then painted using an epoxy, acrylic urethane, or equivalent enamel coating. The paint is selected to be compatible with the pool water and easy to decontaminate.

The paint is visually inspected prior to immersion of the cask in the spent fuel pool and prior to transport. Touch up painting is performed if the paint deteriorates.

2.4.4.3 Lubricants and Cleaning Agents Neolube, Loctite N-5000 or equivalent may be used to coat the threads and bolt shoulders of the TN-40 closure bolts. Never-seez or equivalent is used to coat the contact areas of the top and bottom trunnions prior to lifting operations to prevent impregnation of contamination into the trunnion surface. The lubricant should be selected for compatibility with the spent fuel pool water and the cask materials.

The cask body is cleaned in accordance with approved procedures to remove cleaning residues prior to shipment to the loading site. The basket is also cleaned prior to installation in the cask. The cleaning agents and lubricants have no significant effect on the cask materials and their safety related functions.

Page 2-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.4.4.4 Hydrogen Generation Prairie Islands report to the NRC [13] [14] in response to NRC Bulletin 96-04 demonstrates that galvanic reactions in hydrogen generation are insignificant for the TN-40 cask. Unlike welded canisters, the TN-40 cask has a bolted closure. There is no source of ignition to result in an explosion or fire.

2.4.4.5 Effect of Galvanic Reactions on the Performance of the Cask There are no significant reactions that could reduce the overall integrity of the cask or its contents during storage. The cask and fuel cladding thermal properties are provided in Chapter 3. The emissivity of the basket fuel compartment is 0.3, which is typical for non-polished stainless steel surfaces. If the stainless steel is oxidized, this value would increase, improving heat transfer. The fuel rod emissivity value used is 0.8, which is a typical value for oxidized Zircaloy. Therefore, the passivation reactions would not reduce the thermal properties of the component cask materials or the fuel cladding.

There are no reactions that would cause binding of the mechanical surfaces or the fuel to basket compartment boxes due to galvanic or chemical reactions.

The stainless steel, aluminum, Boral and thermal spray are negligibly affected by the short term exposure to borated water during loading. While formation of blisters in Boral during vacuum drying and heating has been reported, this has not been associated with displacement of the Boral core material containing the boron carbide and therefore has no effect on the Boral criticality safety design function. Furthermore, in the TN-40, the Boral is captured between the structural basket components to provide it with added mechanical support and durability. The outer aluminum lid seals may experience some combination of crevice and galvanic corrosion if they are exposed to water for an extended period of storage prior to transport. However, this would affect only the outer (non-containment) seal, and the seals are tested prior to transport.

There is no significant degradation of any safety components caused directly by the effects of the reactions or by the effects of the reactions combined with the effects of long term exposure of the materials to neutron or gamma radiation, high temperatures, or other possible ambient or operating conditions.

Lifting And Tie-Down Standards Lifting Devices 10CFR 71.45(a) [1] requires that a minimum factor of safety of three against yield is required for all lifting attachments which are structural parts of the package. In addition, the package must be designed such that failure of any lifting device under excessive load would not impair the ability of the package to meet the requirements of 10CFR71

[1]. The stress analyses of the trunnions are provided in the following section.

Page 2-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.5.1.1 Trunnion Analysis The trunnion geometry is shown in Figure 2-1. The front (upper) and rear (lower) trunnions are constructed from SA-105 or SA-266 Class 4 forgings and are groove welded to the cask body. A flat surface is machined on the cask body outer surface at each trunnion location for this purpose.

The following calculation is provided to demonstrate the acceptability of the trunnion design. Additionally, it is shown that the trunnions can be overloaded to failure without compromising the safety of the TN-40 Transport Packaging.

The upper trunnions are used for vertical lifting of the cask. The total weight of the cask is conservatively assumed to be 250,000 lbs. The trunnion is designed to a safety factor of 6 when compared to yield stress and a factor of 10 when compared to ultimate stress. These design loads are very conservative and exceed the 10 CFR 71.45(a) requirements. The load applied to each upper trunnion is then:

FY (UPPER) = 1/2 ( 6 x 250,000 ) = 750,000 lb / trunnion FU (UPPER) = 1/2 ( 10 x 250,000 ) = 1,250,000 lb / trunnion The two lower trunnions are not used for lifting, but only to support the cask as it is upended or downended. Thus they are assumed to support 1/2 of the cask weight since the upper trunnions share the load. Therefore, the load of each of the lower trunnions is:

FY (LOWER) = 1/4 ( 6 x 250,000 ) = 375,000 lb / trunnion FU (LOWER) = 1/4 ( 10 x 250,000 ) = 625,000 lb / trunnion Upper Trunnion Using the dimensions shown in Figure 2-1 the cross sectional area and moment of inertia are:

Section A-A AA A =

4 (12.0 2

)

5.02 = 93.415 in2 I A A =

64 (12.0 4

)

5.04 = 987.20 in4 Page 2-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Shoulder cross section B-B AB B =

4 (11.25 2

)

5.0 2 = 79.73 in 2 IBB =

64 (11.25 4

)

5.0 4 = 755.6 in4 The lifting force applied to the upper trunnion results in a shear load and a bending moment at sections A-A and B-B.

For section A-A, the moment arm, LA-A, is 5.58. For section B-B, the moment arm, LB-B, is 1.75.

For the 6 times lift evaluation, the loads at the respective cross-sections are:

F6x= 750,000 lb M6x A-A= 750,000 lb x 5.58 in. = 4,185,000 in-lb M6x B-B= 750,000 lb x 1.75 in. = 1,312,500 in-lb For the 10 times lift evaluation, the loads at the respective cross-sections are:

F10x= 1,250,000 lb M10x A-A= 1,250,000 lb x 5.58 in. = 6,975,000 in-lb M10x B-B= 1,250,000 lb x 1.75 in. = 2,187,500 in-lb Lower Trunnions Cross sectional area and moment of inertia of the lower trunnions can be calculated using the dimensions provided in Figure 2-1:

Section A-A AA A =

4 (9.5 2

)

4.5 2 = 54.95 in 2 I A A =

64 (9.5 4

)

4.5 4 = 379.69 in 4 Page 2-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Shoulder cross section B-B A BB =

4 (8.88 2

)

4.5 2 = 46.0 in2 IBB =

64 (8.88 4

)

4.5 4 = 285.1 in4 The supporting force applied to the lower trunnion results in a shear load and a bending moment at sections A-A and B-B.

For section A-A, the moment arm, LA-A, is 5.58. For section B-B, the moment arm, LB-B, is 1.75.

For the 6 times lift evaluation, the loads at the respective cross-sections are:

F6x= 375,000 lb M6x A-A= 375,000 lb x 5.58 in. = 2,092,500 in-lb M6x B-B= 375,000 lb x 1.75 in. = 656,250 in-lb For the 10 times lift evaluation, the loads at the respective cross-sections are:

F10x= 625,000 lb M10x A-A= 625,000 lb x 5.58 in. = 3,487,500 in-lb M10x B-B= 625,000 lb x 1.75 in. = 1,093,750 in-lb The section properties and applied loads calculated above are summarized in Table 2-7.

Stress evaluation Upper Trunnions For the 6x lifting load at Section A-A, 750,000 A A = = 8029 psi 93.415 4,185,000 6 B,A A = = 25,436 psi 987 .2 S A A = (25,436) 4(8,029) 2 2 1/ 2

= 30,080 psi For the 6x lifting load at Section B-B, Page 2-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 750,000 BB = = 9,407.0 psi 79.73 1,312,500 5.625 B,B B = = 9,771 psi 755 .6 S B B = (9,771) 4(9,407) 2 2 1/ 2

= 21,200 psi For the 10x support load at Section A-A, 1,250,000 A A = = 13,381 psi 93.5 6,975,000 6 B,A A = = 42,393 psi 987 .2 S A A = (42,393) 4(13,381) 2 2 1/ 2

= 50,134 psi For the 10x support load at Section B-B, 1,250,000 BB = = 15,678 psi 79.73 2,187,500 5.625 B,B B = = 16,285 psi 755 .6 S B B = (16,285) 4(15,678) 2 2 1/ 2

= 35,333 psi Lower Trunnions For the 6x lifting load at Section A-A, 375,000 AA = = 6,824 psi 54.95 2,092,500 B,A A = 4.75 = 26,178 psi 379.69 SAA = (26,178) 4(6,824) 2 2 1/2

= 29,522 psi For the 6x lifting load at Section B-B, 375,000 BB = = 8,152 psi 46.0 Page 2-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 656,250 4.44 B,BB = = 10,220 psi 285.10 SBB = (10,220 ) 4(8,152 )

2 2 1/2

= 19,242 psi For the 10x support load at Section A-A, 625,000 AA = = 11,374 psi 54.95 3,487,500 4.75 B,A A = = 43,629 psi 379.69 SAA = (43,629) 4(11,374) 2 2 1/2

= 49,203 psi For the 10x support load at Section B-B, 625,000 BB = = 13,589 psi 46.0 1,093,750 4.44 B,BB = = 17,033 psi 285.1 S B B = (17,033) 4(13,589) 2 2 1/ 2

= 32,074 psi Table 2-8 presents a summary of the stresses at the same locations to compare against the trunnion yield and ultimate strengths. Also listed are the allowable stresses (yield and ultimate strengths).

The minimum safety margin for the 6 times lift of the upper trunnion is at Section A-A:

M.S. = 31,900/30,080 -1.0 = 0.06 For the lower trunnion, the minimum margin is:

M.S. = 31,900/29,522 -1.0 = 0.08 For the 10 times lift, the minimum safety margins, again at Section A-A are:

M.S. = 70,000/50,134 -1.0 = 0.40 For the lower trunnion, the minimum margin is:

M.S. = 70,000/49,203 -1.0 = 0.43 Page 2-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The results shown above demonstrate that all of the calculated stresses in both the front and rear trunnions are acceptable, and that the minimum margin of safety is 0.06 for the yield condition and 0.40 for the ultimate condition. Both minimums occur in the front trunnions. Therefore the requirements of 10CFR 71.45(a) are met.

10CFR71.45(a) requires that any lifting attachment that is a structural part of the package must not fail in such a manner that the ability of the packaging to meet other requirements is impaired. The trunnions are welded to the gamma shield. The gamma shield is a thick walled cylinder that transmits the lifting load to the balance of the cask.

The Bjilaard analysis provided in Section 2.6 gives a maximum stress in the shell of 20.58 ksi due to the trunnion moment, internal pressure and thermal stress. The outer surface of the gamma shield can be considered the same temperature as the trunnions and thus the allowable stress is identical. Therefore the margin when compared to yield for the shell is much greater than either the trunnion shoulder or weld margin. This ensures the trunnion failure due to excessive load will not affect the performance of the cask because the trunnion will break away from the cask before the cask wall (gamma shield) fails. Note also that the containment vessel is inside the gamma shield and will not be affected by the trunnion failure.

Tie-Down Devices There are no tiedown devices that are a structural part of the package.

The longitudinal forces experienced by the transport package, per 10 CFR 71.45(b), are resisted by steel end restraints which flush up against the impact limiters. The vertical and lateral forces that act on the transport package, according to 10 CFR 71.45(b) and NUREG 766510 [17], are restrained by a dual saddle/strap tie-down system.

Specifically, the tie-down straps resist uplifting and lateral overturning forces whereas the saddles react downward and strap reaction forces. This restraint system is also designed to preclude yielding in the load bearing material of the transport package during normal transport conditions. The premise for both of these tie-down systems is to add extra safety margin by utilizing the large load-bearing surface areas available to distribute transport loads, instead of creating the relatively large localized stresses associated with using the trunnions as transport tie-down points. This loading condition is analyzed in Appendix 2.10.1 (Load Step IL-9). The stress results from the tie-down load are presented in Table 2.10.1-2. All the calculated stresses are less than the lowest yield strength of 31.8 ksi (gamma shield shell).

Page 2-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Normal Conditions Of Transport Overview This section describes the response of the TN-40 package to the loading conditions specified by 10CFR71.71 [1]. The design criteria established for the TN-40 for the NCT are described in Section 2.1.2. These criteria are selected to ensure that the package performance standards specified by 10CFR71.43 and 71.51 [1] are satisfied. Under NCT, there will be no loss or dispersal of radioactive contents, no significant increase in external radiation levels, and no substantial reduction in the effectiveness of the packaging.

Detailed structural analyses of various TN-40 package components subjected to individual loads are provided in the Appendices to this chapter. The limiting results from these analyses are used in this section to quantify package performance in response to the NCT load combinations, specified in 10CFR71.71 [1] and Regulatory Guide 7.8 [6].

Table 2-9 provides an overview of the performance evaluations reported in each load combination subsection. Each subsection provides the limiting structural analysis result for the affected cask component(s) in comparison to the established design criteria.

This comparison permits the minimum margin of safety for a given component subjected to a given loading condition to be readily identified. In all cases, the acceptability of the TN-40 packaging design with respect to established criteria, and consequently with respect to 10CFR71 [1] performance standards is demonstrated.

The structural analysis of the cask body is presented in Appendix 2.10.1 and covers a wide range of individual loading conditions. The stress results from the various individual loads must be combined in order to represent the stress condition in the cask body under the specified condition evaluated in this section. An explanation of the reporting format used for the results, and the stress combination technique used in applying the results from Appendix 2.10.1 is provided here.

Reporting Method for Cask Body Stresses Appendix 2.10.1 provides the detailed description of the structural analyses of the TN-40 cask body. The appendix describes the detailed ANSYS [15] model used to analyze various applied loads. Table 2-10 identifies the individual loads (IL) analyzed which are applicable to NCT.

Detailed stresses are available at as many locations as there are nodes in the finite element model. However, for practical considerations, only the maximum stresses in the lid, flange, inner shell, gamma shield shell, and bottom shield are reported for each load case. These components were selected to be representative of the stress distribution in the cask body. The maximum stress may occur in different components for each individual load.

Page 2-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The stress results for the individual load case (tables reported in Appendix 2.10.1) are for one individual load only. Two or more individual load cases must be combined to determine the total stresses at any stress reporting locations for the various load combinations. This is accomplished using the ANSYS post-processor.

For those load combinations that include trunnion reactions, the local stresses at the trunnion locations found by the Bijlaard [16] method are superimposed on the ANSYS combined stresses.

Table 2-11 provides a matrix of the individual loads, and the various combinations, to determine the cask body stresses for the specified NCT. An x in Table 2-11 indicates that the stress results for the individual load case are used in the load combinations.

For the increased external pressure load combination, it is assumed that the TN-40 cask cavity is at 0 psia. For conservatism, a 25 psig external pressure is used for load combinations.

Heat Chapter 3 describes the thermal analyses of the TN-40 package, subjected to high and low temperature environmental conditions. The analyses results are used to support various aspects of the structural evaluations as described in the following subsections.

2.6.1.1 Maximum Temperatures Allowable stresses for the packaging components are a function of the component temperatures, which are based on actual maximum calculated temperatures or conservatively selected higher temperatures. Chapter 3 summarizes the significant temperatures calculated for the TN-40 package subjected to high temperature environmental conditions. These temperatures are used in establishing the allowable stress values for every NCT load combination, evaluated in this Safety Analysis Report.

Table 2-12 summarizes the thermal analysis results from Chapter 3. The table also lists the selection of cask and basket component design temperatures for structural analysis purposes.

2.6.1.2 Maximum Internal Pressure The thermal analysis, presented in Chapter 3, also provides the average cavity gas temperature under high temperature environmental conditions. This value is used in Chapter 4 to determine the Maximum Normal Operating Pressure (MNOP). For purposes of the structural analysis of containment, a value of 100 psig (much higher than the Chapter 4 value, 15.7 psig) is conservatively assumed for the cask body stress calculation. This pressure loading is analyzed using the ANSYS model of the cask body described in Appendix 2.10.1, and the results are reported in Table 2.10.1-2. This load case and corresponding results are designated as individual load IL-3. IL-3 is used to support evaluations of the load combinations listed in Table 2-11.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.6.1.3 Thermal Stresses (Hot)

The thermal analysis of the TN-40 is performed as described in Chapter 3. The temperature distribution from that analysis is used to perform an ANSYS thermal stress analysis of the cask body. The stress results for this load case are reported on Table 2.10.1-2. This load case is designated as IL-5 (thermal stresses at 100°F ambient) and is used to support various load combinations.

2.6.1.4 Hot Environment Load Combinations (N1)

Cask body stresses for the high temperature environment for NCT, are obtained by a combination of individual loads as summarized in Table 2-11. For this condition, it is assumed that the cask is in its transport configuration, mounted horizontally on the transport cradle, and supported by the front and rear saddles. Pre-load effects on the lid bolts, fabrication stress, 100 psig internal pressure, thermal stresses, and the local stresses at the tiedown straps are combined to give the maximum nodal stress intensity in each component for this load combination. The results are given in Table 2-13 and Table 2-14.

Cold Environment 2.6.2.1 Thermal Stresses for Cold Environment at -20°F Ambient Temperature (N2)

The Regulatory Guide 7.8 [6] requires that the stresses due to the normal load condition to be combined with the thermal stresses for cold environment conditions at -20° F ambient temperature. The thermal stresses are determined in load case IL-6 with results tabulated in Table 2.10.1-2. Again, lid bolt preload, fabrication stress, external pressure, and gravity loads are also included in this combination. The maximum nodal stress intensity in each component for this load combination is listed in Table 2-13 and Table 2-14.

2.6.2.2 Cold Environment Load Combinations at -40°F Ambient Temperature (N3)

The Regulatory Guide 7.8 [6] cold environment load combination results in all cask components in thermal equilibrium at -40 F. Containment vessel thermal stresses do occur in this case due to the differential thermal expansion between the steels. The thermal stresses are determined in load case IL-7 with results tabulated in Table 2.10.1-

2. The cask cavity pressure at the cold environment condition is conservatively assumed to be 0 psia. This results in a net external pressure loading of 14.7 psig (25 psig is conservatively used). The stresses due to 25 psig external pressure are determined in load case IL-4 with results also given in Table 2.10.1-2. Again, lid bolt preload, fabrication stress, and gravity loads are included. The maximum nodal stress intensity in each component for this load combination is listed in Table 2-13 and Table 2-14.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Increased External Pressure (N4)

Cask body stresses for the NCT increased external pressure, 20 psia, are obtained by a combination of individual loads as summarized in Table 2-11. The conservatively assumed minimum cask cavity pressure of 0 psia results in a net external pressure loading of 20 psig (25 psig is conservatively used). For this condition, the cask is assumed to be in the horizontal orientation, supported on the transport cradle front and rear saddles. Lid bolt pre-load, fabrication stress, gravity and the local tiedown strap effects are included. In addition, the thermal stresses for the -20F minimum temperature are also included in the combination. The maximum nodal stress intensity in each component for this load combination is listed in Table 2-13 and Table 2-14.

Reduced External Pressure (N5)

Cask body stresses for the 3.5 psia ambient NCT external pressure decrease are obtained by a combination of individual loads as summarized in Table 2-11. The net internal pressure is calculated as (15.7 + 14.7 - 3.5) = 26.9 psig (cask stresses are conservatively calculated based on 100 psig pressure). For this condition, the cask is in the horizontal orientation supported on the transport cradle by front and rear saddles.

Lid bolt pre-load, fabrication stress, gravity, and the local tiedown strap effects are included. The thermal stresses for the hot thermal condition are included in the load combination. The maximum nodal stress intensity in each component for this load combination is listed in Table 2-13 and Table 2-14.

Transport Shock Loading (N14 & N15)

The transport rail shock loadings used to evaluate the TN-40 transport cask are based on NUREG 766510 [17] which specifies a maximum inertia loading of 4.7g in each of the three x-y-z coordinate directions:

Vertical 4.7g

Longitudinal 4.7g

Lateral 4.7g The resultant transverse load is (4.72 + 4.72)1/2 = 6.65 g The stresses due to the transport rail shock individual load case are presented in Table 2.10.1-2. Table 2-13 and Table 2-14 list the combined stresses (N14) under hot thermal conditions where the load combination is performed for the maximum temperature thermal stresses. Lid bolt pre-load, fabrication stress, internal pressure, and the local tiedown strap effects are included.

In addition, Table 2-13 and Table 2-14 list the combined stresses (N15) under -20°F thermal conditions where the load combination is performed for the -20°F thermal stresses. Lid bolt pre-load, fabrication stress, external pressure, and the local tiedown strap effects are included.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Transport Vibration Loading (N12 & N13)

The input loading conditions used to evaluate the TN-40 cask for transport rail vibration are obtained from NUREG 766510 [17]. The peak inertia values used are:

Vertical 0.37g

Longitudinal 0.19g

Lateral 0.19g The resultant transverse load is (0.372 + 0.192)1/2 = 0.42 g The stresses due to the transport rail car vibration individual load case are presented in Table 2.10.1-2. Table 2-13 and Table 2-14 list the combined stresses (N12) under hot thermal conditions where the load combination is performed for the maximum temperature thermal stresses. Lid bolt pre-load, fabrication stress, internal pressure, and the local tiedown strap effects are included.

In addition, Table 2-13 and Table 2-14 also list the combined stresses (N13) under -

20°F thermal conditions where the load combination is performed for the -20°F thermal stresses. Lid bolt pre-load, fabrication stress, external pressure, and the local tiedown strap effects are included.

Water Spray All exterior surfaces of the TN-40 cask body are metal and therefore not subject to soaking or structural degradation from water absorption. The water spray condition is therefore of no consequence to the TN-40.

Free Drop (N6 through N11)

Two drop orientations are considered credible for the one-foot NCT free drop (see Section 2.10.8.7 of Appendix 2.10.8 for detail descriptions). The structural response of the TN-40 cask body is evaluated for a one-foot end drop of the package on the bottom end, one foot end drop of the package on the lid end, and a one-foot side drop. The assessment of cask body stresses follows the same logic as that established in the previous sections. For the three drop cases, the evaluations are performed for both the hot temperature environment and at the -20F minimum transport temperature.

The load combinations performed to evaluate these drop events are indicated in Table 2-11. In all cases, bolt pre-load effects and fabrication stress are included. For the hot environment condition, thermal stress load, 100 psig internal pressure, and impact load cases are combined. For the cold environment evaluation, -20F thermal stress, 25 psig external pressure, and impact load cases are combined.

Table 2-13 and Table 2-14 list the combined stress intensities for the bottom end, lid end and side drop under hot and cold environment conditions.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Corner Drop This test does not apply to the TN-40 Package since the package weight is in excess of 100 kg (220 lbs.).

Compression This test does not apply to the TN-40 Package since the package weight is in excess of 5,000 kg (11,000 lbs.).

Penetration Due to lack of external protuberances, the one meter (40 inch) drop of a 13 pound steel cylinder of 1-1/4 inch diameter, with a hemispherical head, is of negligible consequence to the TN-40 Package.

Lid Bolt Analysis The lid bolts are analyzed for both NCT and HAC loadings in Appendix 2.10.2. The analysis is based on NUREG/CR-6007 [18]. The bolts are analyzed for the following NCT loadings: operating pre-load, gasket seating load, internal pressure, temperature changes, and impact loads.

The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive ) force on the closure joint, during NCT and HAC events.

A summary of the calculated stresses is listed Section 2.10.2.6. The calculations result in a maximum NCT average tensile stress of 50.1 ksi, which is below the allowable tensile stress of 63.8 ksi. The average NCT shear stress in the bolts is due to torsion during pre-loading. This stress is 13.5 ksi, which is well below the allowable shear stress of 38.3 ksi. The maximum combined stress intensity due to NCT tension plus shear plus bending is 59.3 ksi which is also less than the allowable maximum stress intensity of 86.1 ksi.

The bolt fatigue analysis is also presented in Appendix 2.10.2. This analysis shows that the bolts should be replaced after approximately 50 shipments. This is primarily due to the pre-load stresses.

Fatigue Analysis of the Containment Boundary The purpose of the fatigue analysis is to show that the containment vessel stresses are within acceptable NCT fatigue limits. This is done by determining the fatigue damage factor for each NCT event at locations on the containment vessel with the highest stresses. The cumulative fatigue damage or usage factor for all of the events is conservatively determined by adding the fatigue usage factors for the individual events, assuming these maximum stress intensities occur at the same location.

Page 2-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The fatigue analysis is based on the procedure described in Regulatory Guide 7.6 [5]

and ASME Section III Appendices [7]. When determining the stress cycles, consideration is given to the superposition of individual loads which can occur together and produce a total stress intensity range greater than the stress intensity range of individual loads. Also, the maximum stress intensities for all individual loads are conservatively combined simultaneously. The sequence of events assumed for the fatigue evaluation is given below. The fatigue evaluation is based on 450 shipments.

1. Bolt Preload

2. Lifting

3. Test pressure

4. Road shock/vibration

5. Pressure and temperature fluctuations

6. 1 foot normal condition drop Preload The bolt preload specified to ensure a leak tight seal produces significant stresses in the lid. Therefore, this loading is conservatively included in the fatigue evaluation. The maximum stress calculated in Table 2.10.1-2 is 8,060 psi. It is assumed that the lid is installed twice per round trip resulting in 900 cycles.

Lifting The stresses due to the 6g lifting load are listed in Table 2.10.1-2. The maximum stress intensity, which occurs in the lid, is 2,810 psi. However, when local stress due to the trunnion loading, 15,848 psi calculated in Table 2.10.1-3 is added to the maximum inner shell stress intensity of 1,840 psi (Table 2.10.1-2), the resulting total is 17,688 psi. This value is conservatively used in the fatigue evaluation. This loading is assumed to occur twice per round trip, so the total number of cycles is 900.

Test Pressure The proof test is 1.25 (maximum design pressure) = 125 psi, and will only be performed once. The test pressure stresses are obtained by ratioing the 100 psig internal pressure stresses given in Table 2.10.1-2.

The maximum stress in the flange portion of the containment vessel due to a 100 psi internal pressure is 2,210 psi. Therefore, the stress due to the test pressure is 1.25 x 2,210 = 2,763 psi. This pressure test only occurs once.

Shock Since the TN-40 Cask may be shipped by rail car, the shock and vibration loadings are taken from reference [17].

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Rail Car Shock Rail car shock values were obtained from reference [17]. This reference states that the rail car can be expected to experience a 4.7g load in each direction 9 times every 100 miles. Again, assume 450 round trip shipments, averaging 2000 miles each way.

Therefore the total number of cycles is 2000 (miles) 2 (round trip) 450 (shipments) 0.09 (Shocks per mile) = 162,000 cycles.

The stress intensities due to the rail shock load are listed in Tables 2.10.1-2. The maximum stress intensity in the lid is 1,750 psi.

Vibration According to reference [17], the peak vibration loads at the bed of a railcar are 0.19g longitudinal, 0.19g lateral and 0.37g vertical. The maximum stress intensity resulting from these loads in any of the containment components is 110 psi, which is negligible.

Pressure and Temperature Fluctuations There are four environmental conditions identified for normal condition of transport.

These are hot environment, cold environment, reduced external pressure, and increase external pressure. The containment vessel stresses in response to these environmental load combinations were reported in Table 2-13. The highest total stress intensity from these four cases, 18,090 psi, was calculated to occur in the inner shell during the hot environment condition.

The temperature and pressure fluctuations are assumed to occur once per round trip, since there is no cargo during the return trip, and therefore no pressurization or heat generation. So the total number of cycles of pressure and temperature fluctuation is 450.

Page 2-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 1 Foot NCT Drop The stress intensities due to the 1 foot end drop on bottom are listed in Table 2.10.1-2.

The maximum stress intensity is in the lid portion of the containment vessel and is 120 psi (1g) . Therefore, for a 12 g normal condition end drop, the maximum stress intensity is 120 12 =1,440 psi.

The stress intensities due to the 1 foot end drop on lid end are listed in Table 2.10.1-2.

The maximum stress intensity is in the flange portion of the containment vessel and is 170 psi (1g). Therefore, for 12 g normal condition end drop, the maximum stress intensity is 170 12 = 2,040 psi.

The stress intensities due to the 1 foot side drop are listed in Table 2.10.1-2. The maximum stress intensity at the containment vessel (flange) is 710 psi (1g). Therefore, for 16 g normal condition side drop, the maximum stress intensity is 710 16 = 11,360 psi.

This fatigue evaluation conservatively assumes that the cask is dropped once per shipment, resulting in 450 normal condition drops and using the maximum side stress intensity of 11,360 psi for the damage factor calculation.

Damage Factor Calculation The following table is a summary of the fatigue evaluation. Although the maximum stress intensities for the different loading conditions do not occur at the same location, it is conservatively assumed that they do for the purpose of the fatigue evaluation. The value of the alternating stress, Sa, is determined as follows:

If one cycle goes from 0 to S.I (stress intensity):

Sa = S.I. KF KE /2 If one cycle goes from -S.I. to S.I:

Sa = S.I. KF KE where:

KF = fatigue strength reduction factor, 4 KE = correction factor for modulus of elasticity, 30 106/ 27.8 106 = 1.08 The fatigue curve shown in Table I-9.1 of ASME Section III Appendices [7] is used for this evaluation.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Summary of Fatigue Evaluation Stress Cycles Damage S.I. KF KE Event Intensity Sa (psi) Factor (psi) (psi) n N n/N Lid Stress due 8,060 34,819 17,410 900 165,200 0.01 to Bolt Preload Lifting 17,608 76,412 38,206 900 9,807 0.09 Test Pressure 2,763 11,963 5,968 1 1x106 0.00 Rail Car 1,750 7,560 7,560 162,000 1x106 0.16 Shock Pressure and 18,090 78,150 39,075 450 9205 0.049 Temperature 1 Foot Normal 11,360 49,075 24,538 450 41,000 0.01 Condition Drop 0.319 The above table shows that the total damage factor is less than one. Therefore the fatigue effects on the TN-40 containment vessel are acceptable.

A separate fatigue analysis of the lid bolts is presented in Appendix 2.10.2.

Structural Evaluation of the Basket under Normal Condition Loads The loading conditions considered in the evaluation of the fuel basket consist of inertial loads resulting from NCT drop loading (1 foot drop), HAC drop loading (30 foot drop) and thermal loads. The inertial loads of significance for the basket analysis are those transverse to the cask and basket structural longitudinal axes, so that the loading from the fuel assemblies is applied normal to the basket plates and transferred to the cask wall by the basket.

To determine the structural adequacy of the basket plate in the TN-40 fuel assembly basket under a NCT free drop, the basket is evaluated for 20 g end drop and 20 g side drop. The g-loads and drop orientations used for structural analysis of the basket are described in Appendix 2.10.8. The stress analysis of the basket due to inertial loading analysis is described in detail in Appendix 2.10.5. The results of the analyses are summarized in Appendix 2.10.5, Tables 2.10.5-4 through 2.10.5-6. Based on the results of these analyses, the basket is structurally adequate and it will properly support and position the fuel assemblies under normal loading conditions.

Summary of NCT Cask Body Structural Analysis Table 2-13 lists the highest NCT stress intensities in each of the TN-40 transport package components based on the Section 2.1.2 structural design criteria. From the analysis results presented in Table 2-13 and Table 2-14, it can be seen that the NCT loads will not result in any structural damage to the cask and that the containment function of the cask will be maintained.

Page 2-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Hypothetical Accident Conditions Overview This section describes the response of the TN-40 package to the HAC loading conditions specified by 10CFR71.73 [1]. The design criteria established for the TN-40 packaging for these conditions are described in Section 2.1.2. These criteria are selected to ensure that the packaging performance standards specified by 10CFR71.51 are satisfied.

The presentation of the HAC analyses and results is accomplished in the same manner as that used above for the NCT. The detailed analyses of the various packaging components under different loading conditions are presented in the Appendices to this chapter. The limiting results for the specified HAC loadings are taken from the Appendices and summarized here and compared to the design criteria. In all cases, the acceptability of the TN-40 packaging design with respect to HAC loads is demonstrated.

Table 2-15 provides an overview of the performance evaluations presented in this section. The stress results for the cask body are obtained by combining the stresses from appropriate individual load cases reported in Appendix 2.10.1, to represent the stress condition under the specified HAC. This combination method is essentially the same as that presented in Section 2.6. Stress analysis results for the lid bolts are taken directly from Appendix 2.10.2. The impact limiter attachment evaluations are described in Appendix 2.10.8.

Reporting Method for Cask body Stresses The structural analysis of the cask body was performed using an ANSYS finite element model. Stress results are reported at selected representative locations as described in Section 2.6.

Appendix 2.10.1 provides the detailed description of the structural analyses of the TN-40 cask body. That Appendix describes the detailed ANSYS model used to analyze the cask under various applied loads. Table 2-16 identifies the individual HAC loads (IL) analyzed using the ANSYS model.

Detailed stresses are available at each node in the finite element model. However, for practical considerations, only the maximum stresses in the lid, shell flange, inner shell, shield shell cylinder, and bottom plates are reported for each load case. These components were selected to be representative of the stress distribution in the cask body. The maximum stress may occur in different components for each individual load.

The stress results for the individual load case (tables reported in Appendix 2.10.1) are for one individual load only. Two or more individual load cases must be combined to determine the total stresses at any stress reporting locations for the various load combinations. This is accomplished using the ANSYS post-processor.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 An x in Table 2-17 indicates that the stress results for the individual load case are used in the load combinations.

30 Foot Free Drop In Appendix 2.10.8, the ADOC computer program is used to determine the impact limiter dimensions. The ADOC program is used to estimate the deformation of the impact limiters, the forces on the cask and the cask deceleration due to impact of the packaging on an unyielding surface. The full size impact limiter geometry and wood orientation are designed based on these results.

A one-third scale test impact limiter is fabricated to match the full size impact limiter geometry and wood properties requirements. Four drop orientations are performed to determine the deformations and decelerations of the impact limiters. The test results are used to establish the baseline g loads for the component structural evaluations.

The four drop tests on the one-third scale models of the TN-40 transport package impact limiters are documented in Appendix 2.10.9. For the slapdown drop case, the second impact (combined transverse g load and rotational g load) is a more severe impact to the components than the first impact. Therefore the reported g load for the slapdown is based on the second impact. The maximum g loads for the 90° end drop, 0° side drop, CG over corner drop, and 20° slapdown are as follows:

G Load Measured by Testing 30 Foot Drop Orientation (See Table 2.10.9-1 of Appendix 2.10.9) 90° End Drop 54 g Axial 0° Side Drop 51 g Transverse CG Over Corner Drop 34 g Axial 20° Slapdown (Second Impact) 58 g(1), 62 g(2)

(1) The g load measured at this location represents the maximum combined transverse and rotational g load for the basket structural analysis due to the slapdown drop case.

(2) The maximum combined g load at the top end of the cask body (at the outer surface of the cask lid).

The effect of low temperature (-20 °F) on the tested impact limiter is not available due to lost test data. However, based on the similar design, TN-68 impact limiter testing [30],

chilling the impact limiter wood to -20 °F will increase the g load by 14%. This is based on the measured g load from the TN-68 testing that is approximately 75 g (-20 °F) and the maximum g load predicted by ADOC that is approximately 66 g (room temperature)

(75/66 14%). An increase of 15% in the g loads from testing is used to bound the low temperature effect on the wood properties.

A) Cask Body G Loads Based on the above description, the following table summarizes the baseline g loads for the cask body structural evaluations. First, the g loads are multiplied by the factor due to low temperature effect and then these g loads are increased by additional factors for use as cask bounding baseline g loads.

Page 2-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Baseline G Loads for Cask Body Structural Analyses Bounding Baseline G Loads 30 Foot Bounding Test (-20 °F) Low Used for Cask Body Structural Drop Orientation G Loads Temperature Factor Analyses 90° end drop 54 g axial 1.15 x 54 = 62 68 g (axial)

CG over corner drop 34 g axial 1.15 x 34 = 39 41 g (axial) 0° side drop 51 g transverse 1.15 x 51 = 59 75 g side drop analysis bounds both side 20° slapdown (second drop and slapdown drop impact) 62 g transverse 1.15 x 62 = 71 B) Basket G Loads To establish the baseline g loads, the g loads from the test at basket locations are multiplied by the appropriate dynamic load factors and factors due to low temperature effect (if appropriate).

Baseline G Loads for Basket Structural Analysis Max G Load Low Baseline Ambient Drop Basket Cross from Test DLF Temperature G Load Condition Orientation Section Location (g) Factor(1) Effect(2) (g)

Side drop Mid 51 1.08 - 55 Slapdown Top/Bot 58(3) 1.08 - 63 100 °F End drop Uniform 54 1.08 - 58 Side drop Mid 51 1.08 1.15 63 Slapdown Top/Bot 58(3) 1.08 1.15 72

-20 °F End drop Uniform 54 1.08 1.15 67 (1) Dynamic load factor, see Appendix 2.10.6 (2) Wood property low temperature effect (3) Test g load at basket location C) Fuel Drop G Loads For the fuel drop analyses, the side drop and slapdown orientation, baseline g loads are established by first multiplying the g loads from the test by the appropriate dynamic load factors and factors due to low temperature effect. These g loads are then increased by additional factors and are used as the baseline g loads. Dynamic analysis using testing time history is used in the end drop analysis. The baseline g loads for the fuel drop analyses are listed in the following table.

Baseline G Loads for Fuel Rod Structural Analysis Bounding Test Bounding Baseline G Load Used in Drop Orientation G Loads Load Factor the Fuel Rod Analysis End drop (1) (1) (1)

Side drop 51 g 51 x 1.10 x 1.15 = 65 g (2) (3) 75 g side drop analysis bounds both side Slapdown 58 g 58 x1.10(2) x 1.15(3) = 73 g drop and slapdown drop (1) The acceleration time history curve from the TN40 1/3 scale end drop test (Figure 2.10.9-24) was used. Since the test model was 1/3 of the original size, all of the acceleration values are scaled by 1/3 and all of the times are scaled by a factor of 3. Furthermore, for the -20 °F temperature effect, the acceleration values were increased by 15% and the time values were decreased by 15%.

(2) Dynamic load factor, see Appendix 2.10.7 (3) Wood property low temperature effect Page 2-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 D) Cask Body Structural Analysis The cask body stress evaluations are described in Appendix 2.10.1. The stress analyses were performed before the 1/3 scale impact limiter testing was done; therefore g loads used in Appendix 2.10.1 are based on the ADOC estimation.

Elastic analyses are used for all the cask body drop analyses in Appendix 2.10.1.

Therefore, in order to calculate the stresses due to the bounding baseline g loads resulted from the testing, the load combinations as described in Table 2-17 are performed as follows:

calculated load combination stresses based on the individual load stresses calculated in Appendix 2.10.1

calculated the new load combination stresses by increasing the g values vs.

g values used in the earlier calculations

results of these new load combination stresses are listed in Table 2-18 and Table 2-19 2.7.1.1 End Drop The TN-40 cask body end drop stress analysis performed in Appendix 2.10.1 is based on 1 g. Linear elastic analysis is used for all the cask body structural analysis; the calculated stresses can be ratioed to match the bounding baseline g values. The stress results from the end drop were increased by the factor of baseline g values (68 g) to g values (1 g) used in the calculations.

These increased stress values are used in the end drop load combinations as indicated in Table 2-17 (combination numbers A1 to A4). In all cases, bolt pre-load effects and fabrication stresses are included. For the hot environment condition, 100 psig internal pressure, and impact load cases are combined. For the cold environment evaluation, 25 psig external pressure, and impact load cases are combined.

Table 2-18 lists the maximum nodal combined stress intensities (PL + PB + Q + F) for the bottom and lid end drop under hot environment conditions and cold environment conditions based on the baseline g values.

From Table 2-18, the maximum nodal stress intensity is 21.36 ksi and occurs at the inner shell due to cold load combination for drop on lid end. The membrane allowable is 45.5 ksi (Pm); therefore the minimum factor of safety is 2.13. Note that this stress intensity (21.36 ksi) corresponds to a nodal stress intensity (PL + PB + Q + F) value that is conservatively compared with the Code membrane (Pm) allowable.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.7.1.2 C.G. Over Corner Drop The TN-40 cask body CG over corner drop stress analysis performed in Appendix 2.10.1 is based on the 32 g axial and 14 g transverse. Linear elastic analysis is used for all the cask body structural analysis; the calculated stresses can be ratioed to match the baseline g values. The stress results from the corner drops are increased by a factor of 1.28 applied to earlier calculation g values (32 g) to give baseline g values of 41 g axial and 18 g transverse.

These increased stress values are used in the CG over corner drop load combinations as indicated in Table 2-17 (combination numbers A7 to A10). In all cases, bolt pre-load effects and fabrication stresses are included. For the hot environment condition, 100 psig internal pressure, and impact load cases are combined. For the cold environment evaluation, 25 psig external pressure, and impact load cases are combined.

Table 2-18 and Table 2-19 list the maximum nodal and linearized combined stress intensities for the CG over corner drop under hot environment conditions and cold environment conditions based on the bounding baseline g values.

From Table 2-18, the maximum nodal stress intensity is 28.72 ksi, excluding those marked with an asterisk. This stress occurs at the inner shell due to the cold load combination. The membrane allowable is 45.5 ksi (Pm). Therefore the minimum factor of safety is 1.58. Note that this stress intensity (28.72 ksi) is the nodal stress intensity (PL +

PB +Q +F) value that is conservatively compared with the Code membrane (Pm) allowable.

The stresses marked with an asterisk are linearized and the stresses and factor of safety are listed in Table 2-19. The minimum factor of safety is 1.84.

2.7.1.3 Side Drop/Slapdown The TN-40 cask body side drop stress analysis performed in Appendix 2.10.1 is based on 1 g. Linear elastic analysis is used for all the cask body structural analysis; the calculated stresses can be ratioed to match the baseline g values. The stress results from the side drop were increased by the ratio (75) of baseline g values (75 g) to g values (1 g) used in the calculations. As shown on the cask baseline g load table on Page 2-29, the side drop g load is increased from 59 g to 75 g to bound the slapdown drop cases.

These increased stress values are used in the side drop load/slapdown combinations as indicated in Table 2-17 (combination numbers A5 and A6). In all cases, bolt pre-load effects and fabrication stresses are included. For the hot environment condition, 100 psig internal pressure, and impact load cases are combined. For the cold environment evaluation, 25 psig external pressure, and impact load cases are combined.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 From Table 2-18, the maximum nodal stress intensity is 29.93 ksi, excluding those marked with an asterisk. This stress occurs at the lid due to the hot load combination.

The membrane allowable is 45.5 ksi (Pm); therefore the minimum factor of safety is 1.52. Note that this stress intensity (29.93 ksi) corresponds to a nodal stress intensity (PL + PB +Q +F) value that is conservatively compared with the Code membrane (Pm) allowable.

The stresses marked with an asterisk are linearized and the stresses and factor of safety are listed in Table 2-19. The minimum factor of safety is 1.76.

2.7.1.4 DELETED 2.7.1.5 Lid Bolts The lid bolts are analyzed for normal and accident condition loadings in Appendix 2.10.2. The analysis is based on NUREG/CR-6007 [18]. The bolts are analyzed for the following normal and accident conditions: operating pre-load, gasket seating load, internal pressure, temperature changes, impact loads, and puncture loads.

Page 2-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 A summary of the calculated stresses is listed in Section 2.10.2.6. The calculations result in a maximum HAC average tensile stress of 64.2 ksi, which is below the allowable tensile stress of 87.5 ksi. The average HAC shear stress in the bolts is due to torsion during pre-loading. This stress is 13.5 ksi, which is well below the allowable shear stress of 52.5 ksi.

2.7.1.6 Impact Limiter Attachments The impact limiters must remain attached to the cask body before, during, and after each HAC drop condition.

The limiting loading condition for the impact limiter attachments is the secondary impact (slap-down) associated with the 20 slap down under a 30 foot drop. This loading condition applies the greatest overturning moment to the impact limiter at the cask body interface. Although this loading condition is not limiting with respect to any other cask components, an evaluation of the attachments is performed to demonstrate that the affected impact limiter remains in place to insulate the cask during the subsequent HAC thermal event.

The analysis and results are provided in detail in Section 2.10.8.6.

The analysis concludes that the impact limiter attachment design is sufficiently strong to ensure that the impact limiters remain attached to the cask body during and following all HAC drop events.

Puncture The impact limiters will protect the ends of the cask body from a 40-inch drop onto a 6-inch diameter bar. The most severe damage to the body resulting from the puncture drop will occur on the side walls of the gamma shield shell, between the impact limiters.

This portion of the package is not the containment vessel, so that a release of the contents cannot occur unless both the gamma shield shell and the inner vessel are punctured.

An evaluation of the puncture drop event includes the local effects on the gamma shield shell at the impact point as well as the overall inertia loading on the packaging components.

For this load condition it is assumed that the gamma shield shell surface impacts the puncture bar directly. No credit is taken for the outer shell or the radial neutron shield.

The puncture bar as specified in 10CFR71 [1], is a vertical, cylindrical, mild steel bar 6 inches in diameter.

Page 2-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The impact force exerted by the bar on the gamma shield surface is calculated assuming the bar behaves as an elastic, perfectly plastic material with yield strength of 50 ksi which is a typical yield strength of mild steel. The gamma shield shell is assumed to be SA-266 Cl.4 steel which bounds the three possible fabrication options described in Section 2.1.1 above.

The weight of the TN-40 Transport Package is 271.46 kips. A conservatively higher weight of 275,000 lb is used in this analysis.

Two independent methods are used to compute the stresses in the TN-40 cask shell due to a puncture event.

Puncture Analysis Method 1 The maximum force, Fp, acting on the cask body due to impact on the puncture bar is:

Fp = y Ab Where y is the yield strength of the bar, 50 ksi, and Ab is cross sectional area of the 6 inch diameter bar, 28.27 in.2 Therefore, Fp = 1.414106 lb This force produces a cask deceleration and induces a bending moment at the midsection of the cask. If the cask is considered a beam uniformly loaded (downward) by its inertial load only (conservatively ignoring the 1g gravity force) and supported by the puncture bar at the center, the deceleration g caused by the puncture bar force, Fp, is then the following.

Fp 1.414 10 6 g= = = 5.14 g W package 275,000 If the cask body is considered to be uniformly loaded and supported as described above, then the maximum moment M in the cask shell is:

Fp L (1.414 10 6 )(183.75)

M= = = 3.248 107 in. lb.

8 8 Here, L is the length of the TN-40 cask. Conservatively neglecting the inner shell, outer shell and neutron shield, the moment of inertia of the cask shell is:

I=

4 (r o 4

)

ri 4 =

4 (45.50 4

)

37.5 4 = 1.813 10 6 in.4 Page 2-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The gamma shield shell bending stress is then:

Mr0 (3.248 10 7 )( 45 .50 )

b = = = 815 psi.

I 1.813 10 6 Since the stress is nearly constant through the wall thickness, it should be treated as a membrane stress, Pm. The allowable stress for this accident condition is taken at 300° F as the smaller of 0.7Su (0.7(70,000) = 49,000 psi) or 2.4Sm (2.4(21,300) =

51,120 psi) per Appendix F-1331.1 [7] where SA-105 is the bounding material. The allowable stress of 49,000 psi is well above b.

The thickness of the containment vessel is 8.00 inches which provides the following shear area.

A = (6)(8.00) = 151 in.2 The resulting maximum shear stress is the following.

= Fp/A = 1.414106/151 = 9,364 psi.

The corresponding stress intensity is 2 or 18,750 psi. The allowable stress intensity for the gamma shield (ASME SA-105) is 0.7Su or 0.7(70,000) = 49,000 psi, which is well above the calculated stress intensity.

The deceleration of 5.14 g is small compared to the g-loads that will occur during the 30 foot free drop. Therefore, the global stresses that result from the inertial forces will be neglected during the load combination analysis. The bending stress of 815 psi at the center of the cask is also negligible compared to stresses due to other loads considered.

Puncture Analysis Method 2 An additional cask wall puncture analysis is performed using the equations presented in ORNL NSIC-22 [25]. This method provides a conservative estimate for the puncture threshold thickness of a steel element subjected to nondeformable missile perforation.

The following equation is a problem specific reproduction of the analysis carried out in Reference [25].

(E K ) 2 / 3 Tp =

672 D Where Tp is the cask wall thickness required to prevent puncture, D is the puncture rod outer diameter (in.), and EK is the kinetic energy (ft-lbs) absorbed by the puncture event.

EK is taken to be the potential energy of the TN-40 Transport Package, 40 inches above the puncture rod. Therefore, EK = 275,00040 = 1.100107 in. lb. = 9.167x105 ft. lb.

Substituting Ek and D = 6 in. back into the above equation gives a threshold steel thickness of Page 2-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 (9.167x105 )2 / 3 Tp = = 2.34 in.

(672)(6)

For the purpose of design, the total required thickness is 1.15 (Tp) = 2.69 in. [25].

Since the cask wall is 8.00 inches thick ( > 2.69 inches), the cask wall will not fail due to a puncture event.

Thermal 2.7.3.1 Summary of Pressures and Temperatures The analysis of the thermal accident is presented in Chapter 3. The maximum internal pressure during the HAC thermal accident is calculated in Section 4.3. The calculated pressure is 4.8 atm, or 55.9 psig. The structural analysis is, however, performed conservatively, assuming 100 psig internal pressure for the pressure stress calculations.

An ANSYS transient thermal analysis of the cask for the 30 minute thermal fire accident is reported in Chapter 3. The initial condition is steady state, at an ambient temperature of 100F and maximum decay heat. The initial steady state condition is followed by a 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months fire at 1475°F which is then followed by a cool-down period. The temperatures from the thermal analysis are reported in Chapter 3.

The temperature through the cross section of the cask, at the time of the maximum thermal gradient, is used as input to the cask model for thermal stress analysis.

2.7.3.2 Fire Accident Stresses Stress analysis of the cask body due to fire accident is performed as part of the HAC load combination A15 (Table 2-17. The stress component results of this analysis using the same ANSYS structural model are presented in Table 2-18.

2.7.3.3 Combined Stresses The stress components are combined with those due to the lid bolt pre-load, the internal pressure and fabrication stress, using the same procedure described above for the 30 foot drop events. Table 2-18 and Table 2-19 presents the combined stress intensities in the lid, flange, inner shell, gamma shield shell, bottom, and trunnion region.

Page 2-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Water Immersion 2.7.4.1 Immersion - Fissile Material (Water Head of 3 feet, 1.3 psi External Pressure)

The criticality evaluation presented in Chapter 6 considers the effect of water in-leakage. Thus, the requirements of 10CFR71.73(c)(5) [1] are met. The cask body stresses for this immersion condition (1.3 psi external pressure) is enveloped by the immersion condition for all packages (water pressure of 290 psi) described in Section 2.7.4.3 below.

2.7.4.2 Immersion - All Packages (Water Head of 50 feet, 21.7 psig External Pressure)

The immersion loading condition results in an external pressure applied to the cask body corresponding to a 50 foot head of water. Assuming a 0 psia cask cavity pressure, this results in a maximum external pressure loading of 36.4 psig (21.7 + 14.7).

The cask body stresses resulting from this immersion pressure are enveloped by the immersion condition for all packages (water pressure of 290 psig) described in Section 2.7.4.3 below.

2.7.4.3 Immersion - All Packages (Water Pressure of 290 psig) 10CFR 71.61 [1] requires that the containment vessel be subjected to an external water pressure of 290 psig for a period of not less than one hour without collapse, buckling, or in leakage of water. The containment boundary consists of the inner shell, bottom inner plate shell flange out to the seating surface, and lid assembly outer plate (Figure 4-1).

This analysis evaluates the containment vessel stresses when the 290 psig external pressure is directly applied to the outer surface of the containment vessel. A helpful feature of the packaging design is that the inner shell and bottom inner plate of the containment vessel are completely enclosed by the gamma shield shell and bottom shield. Therefore, they will never be exposed to an external pressure due to immersion.

The finite element model of the cask described in detail in Appendix 2.10.1, is modified to analyze the immersion accident event. The gamma shield structure elements are deleted from the original model. Bilinear material properties were defined for the existing material models to account for plasticity and simulate correct material behavior.

The material properties are obtained from the ASME Code [7]. All properties are taken at 300°F.

All existing loads are removed and replaced by 350 psig pressure loads over the outer surface of the model. The finite element model loads and boundary conditions are shown in Figure 2-2.

A large displacement static analysis is conducted using ANSYS [15]. The 350 psig external pressure is applied in a number of sub-steps. The containment vessel is assumed to buckle at the load sub-step where the solution begins to diverge.

Page 2-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Additionally, results for membrane and bending stresses at the load step of 290 psig external pressure were compared to ASME code allowables.

The critical membrane and membrane plus bending stress intensities for each of the cask components are summarized in the table below. The ultimate strengths of the component materials are obtained from Reference [7].

Containment Vessel Stress Computed Stress Allowable Stress Component Category Intensity (psi) Intensity (psi)

Bottom Inner Plate Pm 21,520 45,500 (section at center) Pm + Pb 47,060 58,500 Inner Plate Pm 8,157 45,500 (bottom plate intersection) Pm + Pb 56,610 58,500 Pm 1,572 49,000 Flange Pm + Pb 4,400 63,000 Lid Outer Plate Pm 4,228 49,000 (section near outer edge) Pm + Pb 6,275 63,000 Lid Shield Plate Pm 836 49,000 (section at center) Pm + Pb 5,856 63,000 A converged solution was obtained for the last sub-step of the analysis corresponding to a 350 psig external pressure, and thus there is no potential of buckling of the containment vessel structure.

Page 2-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The maximum membrane and membrane plus bending stress intensities in each cask component are less than the ASME code HAC allowable stresses. Therefore, it is concluded that cask is adequate for 290 psig external pressure.

In addition to the finite element analysis described above, a buckling evaluation following the methods of ASME Code Case N-284 was performed and is documented in Appendix 2.10.10. This evaluation included the combination of fabrication induced compressive stresses with those due to the 290 psi that result from immersion pressure.

The minimum stress margin is for the combined hoop stress where the amplified plastic stress of 34,748 psi is well below the theoretical buckling stress of 47,657 psi. In addition the interaction check results in a ratio 0.52 which is well below the limit of 1.0.

These results show that the design has significant margins of safety when both fabrication and immersion compressive loads are considered.

Structural Evaluation of the Basket under Accident Loads To determine the structural adequacy of the basket plates in the TN-40 fuel assembly basket under HAC free drops, the basket is conservatively evaluated for a 75 g end drop and a 75 g side drop. The baseline g-loads and drop orientations used for structural analysis of the basket are described in Section 2.7.1. The dynamic load factor used in the basket structural analysis is described in Appendix 2.10.6. The stress analysis of the basket due to inertial loading is described in detail in Appendix 2.10.5.

A summary of the accident analyses performed is presented in Section 2.10.5.6 of Appendix 2.10.5.

The analyses summarized demonstrate that the basket is structurally adequate and will properly support and position the fuel assemblies during HAC loading conditions.

Page 2-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Summary of HAC Cask Body Structural Analysis Table 2-18 lists the highest stress intensities in each of the TN-40 transport package components for all HAC load combinations described above. Also listed in the tables are the stress limits for the service condition based on the Section 2.1.2 structural design criteria.

From the analysis results presented in Table 2-18 and Table 2-19, it can be seen that the HAC loads will not result in any structural damage to the cask and that the containment functions of the cask will be maintained.

As described above, the integrity of the TN-40 Packaging is not compromised by the accident test sequence set forth in 10 CFR71.73 [1], since it meets the design criteria of Regulatory Guide 7.6 [5] for the Load Combinations identified in Regulatory Guide 7.8

[6].

Special Forms / Fuel Rods Special Form This section does not apply to the TN-40 Packaging.

Fuel Rods As discussed in Chapter 4, containment of the radioactive material is provided by the cask containment boundary. Analyses of the cask boundary for NCT and HAC defined by the 10CFR71 [1] demonstrate that the cask remains leak tight.

In addition, Appendix 2.10.7 of the SAR assesses the response of a typical PWR fuel assembly to a 30 foot HAC end drop and a 30 foot HAC side drop. Results from these analyses demonstrate that the fuel rods will not be breached during the NCT and HAC.

Page 2-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. 10 CFR PART 71, Packaging and Transportation of Radioactive Material.

2. Not used.

3. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, 1989.

4. Not used.

5. USNRC Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessel, Rev. 1, March 1978.

6. USNRC Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Cask, Rev. 1, March 1989.

7. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Appendices, 1989.

8. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section V, 1989.

9. Not used.

10. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section IX, 1989.

11. NUREG/CR-3854,Fabrication Criteria for Shipping Container.

12. NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, March 1998.

13. Northern States Power Company, Response to Bulletin 96-04, August 19, 1996, Docket No. 72-10, Materials License No. SNM-2506.

14. Hydrogen Generation Analysis Report for TN-40 Cask Materials, Test Report No.

61123-99N, Rev 0, Oct 23, 1998, National Technical Systems.

15. ANSYS Computer Code and Users Manuals, Release. 8.0.

16. WRC Bulletin 107, March 1979 Revision Local Stresses in Spherical and Cylindrical Shells Due to External Loadings.

17. NUREG 766510, Shock and Vibration Environments for Large Shipping Containers on Rail Cars and Trucks, June, 1977.

18. NUREG/CR-6007, "Stress Analysis of Closure Bolts for Shipping Casks", Lawrence Livermore National Laboratory, 1992.

Page 2-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

19. Not used.

20. Not used.

21. NUREG/CR-3826, Recommendations for Protecting against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, April 1984.

22. NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers up to Four Inches Thick.

23. Aluminum Standards and Data-1, 1976, published by the aluminum association.

24. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section VIII, Divisions 1 and 2, 1989.

25. Gwaltney, R.C., Missile Generation and Protection in Light-Water-Cooled Power Reactor Plants, ORNL NSIC-22, Oak Ridge National Laboratory, Oak Ridge, TN, For the UXAEC, September 1968.

26. Not used.

27. Not used.

28. Transnuclear ADOC computer program, Rev. 1.

29. Not used.

30. TN-68 Transport Packaging Safety Analysis Report, Revision 3 (CoC 9293, Docket No. 71-9293).

Page 2-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Appendices The detailed structural analyses of the TN-40 packaging are included in the following appendices:

Appendix 2.10.1 Structural Analysis of the Cask Body Appendix 2.10.2 Lid Bolt Analysis Appendix 2.10.3 Structural Analysis of the Outer Shell Appendix 2.10.4 Fracture Toughness Evaluation of the TN-40 Cask Appendix 2.10.5 Structural Analysis of the TN-40 Basket Appendix 2.10.6 Dynamic Load Factor for Basket Drop Analysis Appendix 2.10.7 Structural Evaluation of the Fuel Rod Cladding Under Accident Impact Appendix 2.10.8 Structural Evaluation of the Impact Limiters Appendix 2.10.9 TN-40 Impact Limiter Testing Appendix 2.10.10 Inner Shell Buckling Due to External Pressure Appendix 2.10.11 TN-40 Lid Closure Evaluation Due to Delayed Impact Page 2-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 ASME Code Alternatives Both the cask containment boundary and basket are designed, fabricated and inspected in accordance with the ASME Code Subsection NB to the maximum practical extent. The gamma shielding, which is primarily for shielding, but also provides structural support to the containment boundary during NCT and HAC events, was designed in accordance with Subsection NF of the code. Inspections of the gamma shielding are performed in accordance with ASME code Subsection NF as detailed in the SAR. Other cask components, such as the protective cover, outer shell and neutron shielding are not governed by the ASME Code.

Reference ASME Component Code Requirement Alternatives, Justification & Compensatory Measures Code/Section The TN-40 cask is not N/TP stamped, nor is there a code design specification or stress report generated. A design Stamping and criteria document is generated in accordance with preparation of reports NB-1100/ Subsection Transnuclears QA Program and the design and analysis is by the Certificate TN-40 Cask NCA performed under TNs QA Program and presented in the SAR.

Holder, Surveillances, NB-2000 The cask may also be fabricated by other than N-stamp holders Use of ASME and materials may be supplied by other than ASME Certificate Certificate Holders holders. Surveillances are performed by TN and utility personnel rather than by an Authorized Nuclear Inspector (ANI)

The quality assurance requirements of NQA-1 or 10 CFR 71 are TN-40 Cask NCA-3800 QA Requirements imposed in lieu of NCA-3800 requirements.

The containment vessel is hydrostatically tested in accordance with the requirements of the ASME B&PV Code,Section III, Pressure Test of the Articles NB-6200 with the exception that some of the NB-6200 Hydrostatic Testing Containment Boundary containment vessel is installed in the gamma shield shell during testing. The containment vessel is supported by the gamma shield during all design and accident events.

Full penetration The required UT inspection will be performed on a best efforts corner welded joints basis. The joint will be examined by RT and either PT or MT Weld of bottom inner require the fusion methods in accordance with ASME Subsection NB plate to the containment NB-5231 zone and the parent requirements. The joint may be welded after the containment shell metal beneath the shell is shrink fitted into the gamma shield shell. The geometry attachment surface to of the joint does not allow for UT inspection.

be UT after welding.

Page 2-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Reference ASME Component Code Requirement Alternatives, Justification & Compensatory Measures Code/Section The rolling process used to form the inner vessel should be qualified to determine that the required Containment Shell If the plates are made from less than three heats, each heat will NB-4213 impact properties of Rolling Qualification be tested to verify the impact properties.

NB-2300 are met after straining by taking test specimens from three different heats.

No overpressure protection is provided. Function of containment vessel is to contain radioactive contents under Vessels are required normal and accident conditions of transport. Containment Containment Vessel NB-7000 to have overpressure vessel is designed to withstand maximum internal pressure protection considering 100% fuel rod failure and maximum accident temperatures.

Requirements for TN-40 cask is to be marked and identified in accordance with nameplates, stamping 10 CFR71 requirements. Code stamping is not required. QA Containment Vessel NB-8000 and reports per NCA- data package to be in accordance with TN approved QA 8000 program.

The design specification shall define the boundary A code design specification was not prepared for the TN-40 Containment Vessel NB-1131 of a component to cask. A TN design criteria specification was prepared in which other accordance with TNs QA program.

component is attached.

Page 2-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Reference ASME Component Code Requirement Alternatives, Justification & Compensatory Measures Code/Section Standard Review Plan, NUREG-1536 has accepted the use of either Subsection NB (Class 1) or NC (Class 2 or 3) of the Code for the containment. SA-203 Grade. D is similar to SA-203 Grade E which is a Class 1 material. The chemical content of the two grades are identical, except that Grade E restricts the carbon to 0.20 max., while Grade D further restricts the carbon content to 0.17 max. Grade. D is acceptable as a class 2 material up to 500°F.

Containment Vessel Materials to be ASME NB-2120 Grade D was selected because of its ductility, since the higher Material Class 1 material strength is not required. SA-203 Grade D has better elongation than Grade E and due to its lower strength is more likely to have the good fracture toughness at low temperatures.

In selecting materials for storage and transport casks, one of the major selection criteria is fracture toughness at low temperatures. Grade D was selected on this basis. There is no similar requirement for pressure vessels, as they are used at much higher temperatures.

If two different materials are joined, the fracture toughness requirements of either may be used for the weld metal. There Impact testing of weld are no fracture toughness requirements on the shield plate, and Weld of Shield Plate to and heat affected NB-4335 therefore none are performed on the base metal or the heat Lid Outer Plate zone of lid to shield affected zones. This weld is not subject to low temperatures, as plate it is inside the cask cavity. An evaluation of this weld at low temperatures is presented in Appendix 2.10.4 of the SAR.

Requires materials to Material will be supplied by TN approved suppliers with Certified be supplied by ASME Containment Vessel and Material Test reports (CMTR) in accordance with NB-2000 approved material Lid Penetration Cover NB-2000 requirements. The cask is not code stamped. The quality supplier; Quality Materials assurance requirements of NQA-1 or 10CFR71 are imposed in assurance to meet lieu of the requirements of NCA-3800.

NCA requirements.

Non-pressure The primary function of the gamma shield is shielding, although retaining structural credit is taken for the gamma shielding in the structural Gamma Shielding NB-1132.2 attachments shall analysis. The welds are examined in accordance with NF conform to acceptance criteria. A fracture toughness evaluation is Subsection NF. presented in Appendix 2.10.4 of the SAR.

Page 2-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Reference ASME Component Code Requirement Alternatives, Justification & Compensatory Measures Code/Section Material in the component support The gamma shielding materials were procured to ASTM or load path and not ASME material specifications. Materials testing is performed in performing a pressure accordance with the applicable specification. Impact testing is retaining function Gamma Shielding NB-2190 not performed on the gamma shielding materials (including welded to pressure welding materials). An evaluation of the gamma shielding due retaining material to impact at low temperatures is provided in Appendix 2.10.4 of shall meet the the SAR.

requirements of NF-2000.

Repetition of surface Lid & Flange NB-4121.3 examination after Critical Flaw size determination is performed in Appendix 2.10.4 machining.

Basket fabrication and welding procedures are qualified in Fabrication/Welding / accordance with ASME Section IX. Due to the unique nature of Basket NB-4000/5000 NDE inspection these basket and welds, special inspections and tests were developed for these welds.

The basket neutron poison material is not used for structural Basket neutron poison Use of ASME NB-2000 analysis, but to provide criticality control and heat transfer.

material Materials They are not code materials.

The aluminum plate is not a Class 1 material. It was selected for its properties. Aluminum has excellent thermal conductivity and a high strength to weight ratio. NUREG-3854 and 1617 Aluminum used for allow materials other than ASME Code materials to be used in basket rails, aluminum the cask and basket fabrication. ASME Code does provide the plates between the Use of ASME material properties for the aluminum and also allows the NB-2000 compartments, and Materials material to be used for Section III applications (Class 2 and 3).

aluminum plates at the Note: Sm and S values for aluminum are taken from ASME basket periphery Code Section VIII, Division 2. S is taken from Section III or the lower of Su/4 or 2/3 Sy. Aluminum material properties at elevated temperatures are taken from Aluminum Standards and Data [23].

Page 2-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Reference ASME Component Code Requirement Alternatives, Justification & Compensatory Measures Code/Section Factors of safety are used to account for geometrical imperfections, residual stresses, load eccentricity, and modeling assumptions. The buckling analyses reported in Section 2.10.5.3 have a minimum safety factor of 1.18 (88.54 / 75). The analyses modeled a full 360 degree sector of the basket with elastic-plastic material and large deflection effects; in addition the analyses also used a conservative value for the fuel weight.

ASME Section III, Applied load shall not Sensitivity analyses were performed (Section 2.10.5.5) to Basket buckling analysis Appendix F-1341.4 exceed 0.7 PL investigate the effect of geometrical imperfections and allowing the basket cell walls to be able to buckle away from each other.

The results showed that the minimum safety factor of 1.18 is still bounding. Furthermore, compression tests were performed to determine the basket panel strength during side impact (Section 2.10.5.5.3). The results showed a safety factor of 1.73 with respect to the design load of 75g. These analyses show the basket to be structurally adequate.

Page 2-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-1 Evaluation Method Employed to Demonstrate Compliance With Specific Regulatory Requirements Numerical Material Model 10CFR71 Analysis Test** Tests Heat X Cold X Normal Reduced External Pressure X Condition Increased External Pressure X of Transport Shock and Vibration X One Foot Free drop X 30 foot Free Drop - Cask and Basket X X Hypothetical 30 foot Free Drop- Impact Limiters X X X Accident Puncture X Condition Thermal Event X Water Immersion X Lifting X others Tie-Down X

- Material tests include crush and shear tests of the wood, and charpy and tensile tests of the containment materials.

Page 2-49

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-2 Containment Vessel Stress Limits Classification Stress Intensity Limit Normal (Level A) Conditions(1)

Pm Sm Pl 1.5 Sm (Pm or Pl) + Pb 1.5 Sm Shear Stress 0.6 Sm Bearing Stress Sy (Pm or Pl) + Pb + Q 3 Sm (Pm or Pl) + Pb + Q + F Sa Hypothetical Accident (Level D)(2)

Pm 0.7 Su Pl Su(3)

(Pm or Pl) + Pb Su(3)

Shear Stress 0.42 Su Notes:

1. Classifications and Stress Intensity Limits are as defined in ASME B&PV Code,Section III, Subsection NB [3].

2. Stress intensity limits are in accordance with ASME B&PV Code,Section III, Appendix F [7].

3. When evaluating the results from the nonlinear elastic plastic analysis for the accident conditions, the general primary membrane stress intensity, Pm, shall not exceed 0.7Su and maximum primary stress intensity at any location (Pl or Pl + Pb) shall not exceed 0.9Su [7].

Page 2-50

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-3 Cover Bolt Stress Limits Classification Stress Intensity Limit(1)(5)

Normal (Level A) Conditions (2)

Average Tensile Stress 2 Sm Maximum Combined Stress 3 Sm Bearing Stress Sy Hypothetical Accident (Level D) (3)

Average Tensile Stress Smaller of Sy or 0.7 Su Average Shear Stress Smaller of 0.42 Su or 0.6 Sy Maximum Combined Stress Su Combined Shear & Tension (4)

Rt + Rs2 1 2

Notes:

1. The stress analysis of the lid bolts is performed in accordance with NUREG/CR-6007 [18]

described in Appendix 2.10.2. The stress limits for the lid bolts are listed separately in Appendix 2.10.2, Tables 2.10.2-3 and 2.10.2-4.

The lid bolt evaluation due to delayed impact is presented in Appendix 2.10.11 using LS-DYNA analysis; the stress in the bolt is limited to the yield strength of the lid bolt material.

The stress limits for the impact limiter tie rod and attachment bolt are described in Appendix 2.10.8 Structural Analysis of Impact Limiter.

2. Classification and stress limits are as defined in ASME B&PV Code,Section III, Subsection NB

[3].

3. Stress limits are in accordance with ASME B&PV Code,Section III, Appendix F [7].

4. Rt : Ratio of average tensile stress to allowable average tensile stress Rs : Ratio of average shear stress to allowable average shear stress

5. All stresses include the effect of tensile and torsional loads due to bolt preloading.

Page 2-51

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-4 Non Containment Structure Stress Limits Classification Stress Intensity Limit Normal (Level A) Conditions (1)

Pm Sm Pl 1.5 Sm (Pm or Pl) + Pb 1.5 Sm (Pm or Pl) + Pb + Q 3 Sm Shear Stress 0.60 Sm Bearing Stress Sy Hypothetical Accident (Level D) (2)

Pm 0.7 Su Pl SU (Pm or Pl) + Pb SU Shear Stress 0.42 Su Weld Allowable(1)

Normal Load Condition Full Penetration Same as base metal Partial Grove/Fillet Tension - 0.3 x Su Shear - 0.4 x Sy Accident Load Condition Full Penetration Same as base metal Partial Grove/Fillet Normal Condition allowables are increased by a factor: Smaller of 2 or 1.67Su/Sy if Su > 1.2Sy Notes:

1. Classifications and stress intensity limits are as defined in ASME B&PV Code,Section III, Subsection NF [3].

2. Stress intensity limits are in accordance with ASME B&PV Code,Section III, Appendix F [7].

Page 2-52

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-5 Basket Stress Limits Classification Stress Intensity Limit Normal (Level A) Conditions (1)

Pm Sm Pl 1.5 Sm (Pm or Pl) + Pb 1.5 Sm (Pm or Pl) + Pb + Q 3 Sm (Pm or Pl) + Pb + Q + F Sa Hypothetical Accident (Level D) (2)

Pm Smaller of 2.4 Sm or 0.7 Su Pl Smaller of 3.6Sm or Su(3)

(Pm or Pl) + Pb Smaller of 3.6 Sm or Su(3)

Notes:

1. Classifications and stress intensity limits are as defined in ASME B&PV Code,Section III, Subsection NB [3].

2. Limits are in accordance with ASME B&PV Code,Section III, Appendix F [7].

3. When evaluating the results from the nonlinear elastic plastic analysis for the accident conditions, the general primary membrane stress intensity, Pm, shall not exceed 0.7Su and maximum primary stress intensity at any location (Pl or Pl + Pb) shall not exceed 0.9Su [7].

4. Fusion welds are qualified by testing. The testing program is provided in Section 2.10.5.4.

Page 2-53

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-6 Cask Weight and Center of Gravity Component Weight (kips)

Cask Body 116.33 Lid 13.91 Bottom 18.87 Aluminum Boxes 1.99 Resin 10.58 Outer Shell 7.45 Trunnions 0.67 Fuel Assemblies 52.00 Basket Steel Boxes 5.44 Aluminum Plates of the Basket 5.40 Boral Plates 0.58 Steel Plates at Periphery 1.17 Aluminum Rails at Periphery 1.29 Aluminum Plates at Periphery 0.81 Top Impact Limiter 16.34 Bottom Impact Limiter 16.33 Tie Rods (13) 0.91 Impact Limiter Bolting Brackets (8) 0.24 Top Impact Limiter Spacer 1.15 Total with Fuel, Impact Limiters and Tie Rods 271.46

Center of Gravity of the package is approximately 91.4 inches and is measured along the axial centerline from the rear (bottom) of the cask.

Summary of weights used for Analysis:

1. Front (Top) Trunnion Lifting (w/o impact limiters) 250,000 lbs.

2. Cask Body Analysis 271,700 lbs.

Page 2-54

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-7 Trunnion Section Properties and Applied Loads Front Trunnions Rear Trunnions Item Section Section Section Section A-A B-B A-A B-B (Weld) (Shoulder) (Weld) (Shoulder)

Cross Section Area (in2) 93.415 79.73 54.95 46.0 Area Moment Of Inertia (in4) 987.20 755.60 379.69 285.10 Yield Condition*

750,000 375,000 Shear Force (lb)

Yield Condition*

4,185,000 1,312,500 2,092,500 656,250 Bending Moment, (in lb)

Ultimate Condition**

1,250,000 625,000 Shear Force (lb)

Ultimate Condition**

6,975,000 2,187,500 3,487,500 1,093,750 Bending Moment (in lb)

Notes:

Trunnion geometry (Sections A-A and B-B) is in Figure 2-1.

Trunnion Loads to Support 6 times Cask Weight

- Trunnion Loads to Support 10 times Cask Weight Page 2-55

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-8 Trunnion Stresses when Loaded by 6 and 10 Times Cask Weight (Lifting)

Location / Stress Yield Limit Ultimate Limit Section Section Section Section A-A B-B A-A B-B (Weld) (Shoulder) (Weld) (Shoulder)

Front Trunnions Shear Stress (psi) 8,029 9,407 13,381 15,678 Bending Stress (psi) 25,436 9,771 42,393 16,285 Stress Intensity (psi) 30,080 21,200 50,134 35,333 Rear Trunnions Shear Stress (psi) 6,824 8,152 11,374 13,589 Bending Stress (psi) 26,178 10,220 43,629 17,033 Stress Intensity (psi) 29,522 19,242 49,203 32,074 Allowable Stress (psi) Sy = 31,900 Su = 70,000 Notes:

1. Trunnion geometry (Sections A-A and B-B) is shown in Figure 2-1.

2. Minimum margin of safety is 0.06 for yield limit and 0.40 for ultimate limit (front trunnion).

Page 2-56

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-9 TN-40 Performance Evaluation Overview (Normal Conditions of Transport)

Loading Condition SAR Section Scope of Evaluation Maximum component temperatures for material 2.6.1.1 allowables Heat 2.6.1.2 Cask cavity maximum pressure, 100 psig 71.71(c)(1) 2.6.1.3 Cask body thermal gradients Cask body stresses due to hot environment load 2.6.1.4 combinations Cold Cask body stresses due to cold environment load 2.6.2 71.71(c)(2) combinations Increase External Cask body stresses due to 25 psig external pressure load Pressure 2.6.3 combinations 71.71(c)(4)

Reduced External Cask body stresses due to 100 psig internal pressure pressure 2.6.4 load combinations 71.71(c)(3)

Shock Loads 2.6.5 Cask body stresses due to rail shock loads 71.71(c)(5)

Vibration Loads 2.6.6 Cask body stresses due to rail vibration loads 71.71(c)(5)

Water Spray 2.6.7 Negligible for TN-40 cask 71.71(c)(6)

Cask body stresses due to 1 foot bottom end drop Free Drop 2.6.8 Cask body stresses due to 1 foot lid end drop 71.71(c)(7)

Cask body stresses due to 1 foot side drop Corner Drop 2.6.9 Not applicable 71.71(c)(8)

Compression 2.6.10 Not applicable 71.71(c)(9)

Penetration 2.6.11 Not applicable 71.71(c)(10)

Bolt stresses due to preload, pressure loads, Lid Bolt Analysis 2.6.12 temperature, impact and puncture loads Fatigue evaluation of containment vessel due to lifting, Fatigue Analysis of 2.6.13 pressure, temperature, shock/vibration, and 1 foot drop Containment Boundary loads Structural analysis of the basket due to 1 foot end drop Basket Evaluation 2.6.14 and 1 foot side drop loads Summary of Normal Lists the highest stress intensities in the containment Condition Structural 2.6.15 vessel and gamma shield and compares results with the Analysis allowables Page 2-57

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-10 Individual Load Cases for Normal Conditions of Transport TN-40 Cask Body Analysis Factor Used for Load used in Run No. Individual Load Type Normal Conditions Analysis Load Combinations IL-1 Bolt preload and lid seating pressure - 1.0 IL-2 Fabrication Stresses - 1.0 IL-3 Internal pressure 100 psig 1.0 IL-4 External pressure 25 psig 1.0 Thermal stresses at 100°F hot IL-5 - 1.0 environment Thermal stresses at -20°F cold IL-6 - 1.0 environment Thermal stresses at -40°F cold IL-7 - 1.0 environment IL-8 Cask Horizontal - 1g Down Gravity 1g 1.0 Horizontal cask, Rail Vibrations IL-10 0.19g,0.19g, 0.37g 1.0 Rail car shock Horizontal cask, Rail Shock IL-11 4.7g all direction 1.0 tie-down IL-12 End drop on bottom 1g 12.0 IL-13 End drop on lid 1g 12.0 IL-15 Side drop 1g 16.0 Page 2-58

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-11 Summary of Load Combinations for Normal Condition of Transport (2 Sheets)

Applicable Individual Loads Load IL-1 IL-2 IL-8 IL-3 IL-4 IL-5 IL-6 IL-7 IL-12 IL-13 IL-15 Combination Comb.

Bolt Gravity Internal External Themal Thermal Thermal Bottom Top Side Fabrication Abbr.

Pre-load 1g Pressure Pressure 100°F -20°F -40°F Drop Drop Drop Hot Environment X X X X X N1 (100° F amb.)

Cold Environment X X X X X N2

(-20° F amb.)

Cold Environment X X X X X N3

(-40° F amb.)

Increased External X X X X X N4 Pressure Reduced External X X X X X N5 Pressure 1 Ft End X X X X X N6 Bottom Drop X X X X X N7 1 Ft End Top X X X X X N8 Drop X X X X X N9 X X X X X N10 1 Ft Side Drop X X X X X N11 Page 2-59

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-11 Summary of Load Combinations for Normal Condition of Transport (2 Sheets)

Applicable Individual Loads Load IL-1 IL-2 IL-3 IL-4 IL-5 IL-6 IL-10 IL-11 Combination Combination Bolt Internal External Thermal Thermal Rail Rail Fabrication Abbreviation Pre-load Pressure Pressure (100°F) (-20°F) Vibration Shock X X X X X N12 Rail Vibration X X X X X N13 X X X X X N14 Rail Shock X X X X X N15 Page 2-60

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-12 Reference Temperatures for Stress Analysis Acceptance Criteria Normal Transport*

Maximum from Selected Design**

Component Chapter 3 Temperature (F) (°F)

Outer Shell 214 250 Inner Shell 251 300 Basket Rail 257 ***

Basket Plate 444 ***

Gamma Shell 248 300 Fuel Cladding 495 500 Lid Bolt <250 300

For normal loading condition.

- Temperatures specified are used to determine allowable stresses.

They are not a maximum use temperature for material.

- - Allowable stresses for the basket are taken at the temperatures.

Page 2-61

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-13 Summary of Load Combination Stresses for Normal Conditions of Transport (2 Sheets)

Cask Component Nodal Stress Intensity (ksi)

Load Stress Gamma Allowable Comb. Type Inner Bottom (ksi)

Lid Flange Shield Trunnion Shell Shield Shell N1 Primary 8.70 12.65 13.74* 7.32 7.13 3.70 19.6 Hot Primary +

9.17 14.93 18.09 7.29 15.50 3.84 58.8 (100° F) Secondary N2 Primary 8.19 13.38* 14.63* 5.72 3.32 3.10 19.6 Cold Primary +

(-20° F) 8.27 14.52 17.46 5.30 7.19 3.19 58.8 Secondary N3 Primary 8.19 13.38* 14.63* 5.72 3.32 3.10 19.6 Cold Primary +

(-40° F) 8.22 14.29 17.15 5.27 6.50 3.16 58.8 Secondary N4 Primary 8.19 13.38* 14.63* 5.72 3.32 3.10 19.6 External Primary +

Pressure 8.27 14.52 17.46 5.30 7.19 3.19 58.8 Secondary N5 Primary 8.70 12.65 13.74* 7.32 7.13 3.70 19.6 Internal Primary +

Pressure 9.17 14.93 18.09 7.29 15.50 3.84 58.8 Secondary N6 Primary 8.10 13.88* 15.72* 4.65 3.50 3.11 19.6 Drop Bottom Primary +

(C) 8.19 15.03 18.56 4.31 7.26 3.29 58.8 Secondary N7 Primary 8.59 13.13* 14.84* 6.23 7.15 3.70 19.6 Drop Bottom Primary +

(H) 9.06 15.43 19.19 6.27 15.62 3.83 58.8 Secondary N8 Primary 8.93 14.35* 15.03* 4.86 3.36 3.18 19.6 Drop Top Primary +

(C) 9.00 15.52 17.88 4.41 5.21 3.72 58.8 Secondary N9 Primary 9.41 13.60* 14.13* 6.41 5.00 3.67 19.6 Drop Top Primary +

(H) 9.88 15.91 18.50 6.36 13.26 4.11 58.8 Secondary N10 Primary 9.55 13.40* 14.74* 13.49* 7.93 4.26 19.6 Drop Side Primary +

(C) 9.67 14.66 17.56 13.39 10.97 4.20 58.8 Secondary N11 Primary 9.86 12.67 13.85* 13.44* 12.78 4.82 19.6 Drop Side Primary +

(H) 10.19 15.07 18.17 13.38 19.31 4.89 58.8 Secondary Page 2-62

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-13 Summary of Load Combination Stresses for Normal Conditions of Transport (2 Sheets)

Cask Component Nodal Stress Intensity (ksi)

Load Stress Gamma Allowable Comb. Type Inner Bottom (ksi)

Lid Flange Shield Trunnion Shell Shield Shell Primary 8.70 12.66 13.75* 7.24 7.02 3.69 19.6 N12 Vib. (H) Primary +

9.17 14.94 18.11 7.23 15.43 3.83 58.8 Secondary Primary 8.20 13.39* 14.64* 5.64 3.30 3.09 19.6 N13 Vib. (C) Primary +

8.27 14.53 17.48 5.25 7.12 3.18 58.8 Secondary Primary 8.64 12.83 14.19* 7.52 8.24 3.81 19.6 N14 Shock (H) Primary +

9.05 15.15 18.53 7.30 16.27 4.21 58.8 Secondary Primary 8.23 13.56* 15.08* 5.91 3.84 3.29 19.6 N15 Shock (C) Primary +

8.31 14.74 17.91 5.28 7.91 3.58 58.8 Secondary

The stresses which result in factor of safety less than 1.5, are linearized in Table 2-14.

Page 2-63

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-14 Linearized Stress Evaluation of Normal Condition of Transport Load Combinations (4 Sheets)

Nodal Linearized Stress Intensity Stress Factor Load Intensity Node Magnitude Allowable of Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety N1 (Hot) Inner Shell 13.74 224-254 PM 13.01 19.6 1.51 PL + PB 13.47 29.4 2.18 N2 (Cold) Flange 13.38 4482- PM 4.90 19.6 4.00 4485 PL + PB 11.88 29.4 2.47 Inner Shell 14.63 373- PM 13.63 19.6 1.44 15591 PL + PB 14.06 29.4 2.09 N3(Cold) Flange 13.38 4482- PM 4.90 19.6 4.00 4485 PL + PB 11.88 29.4 2.47 Inner Shell 14.63 373- PM 13.63 19.6 1.44 15591 PL + PB 14.06 29.4 2.09 N4(Cold) Flange 13.38 4482- PM 4.90 19.6 4.00 4485 PL + PB 11.88 29.4 2.47 Inner Shell 14.63 373- PM 13.63 19.6 1.44 15591 PL + PB 14.06 29.4 2.09 N5 (Hot) Inner Shell 13.74 224-254 PM 13.01 19.6 1.51 PL + PB 13.47 29.4 2.18 Page 2-64

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-14 Linearized Stress Evaluation of Normal Condition of Transport Load Combinations (4 Sheets)

Nodal Linearized Stress Intensity Stress Factor Load Intensity Node Magnitude Allowable of Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety N6 (Cold) Flange 13.88 4482- PM 5.07 19.6 3.87 Bottom 4485 PL + PB 12.25 29.4 2.40 Drop Inner Shell 15.72 224-254 PM 13.52 19.6 1.45 PL + PB 13.98 29.4 2.10 N7(Hot) Flange 13.13 375-374 PM 4.77 19.6 4.11 Bottom PL + PB 11.71 29.4 2.51 Drop Inner Shell 14.84 224-254 PM 13.00 19.6 1.51 PL + PB 13.46 29.4 2.18 N8(Cold) Flange 14.35 375-374 PM 5.54 19.6 3.54 Top Drop PL + PB 12.66 29.4 2.32 Inner Shell 15.03 373- PM 13.6 19.6 1.44 15591 PL + PB 14.02 29.4 2.10 N9(Hot) Flange 13.60 375-374 PM 5.23 19.6 3.75 Top Drop PL + PB 12.12 29.4 2.43 Inner Shell 14.13 373- PM 13.03 19.6 1.50 15591 PL + PB 13.44 29.4 2.19 Page 2-65

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-14 Linearized Stress Evaluation of Normal Condition of Transport Load Combinations (4 Sheets)

Nodal Linearized Stress Intensity Stress Factor Load Intensity Node Magnitude Allowable of Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety N10 Flange 13.40 4655- PM 5.22 19.6 3.75 (Cold) 4654 PL + PB 11.92 29.4 2.47 Side Drop Inner Shell 14.74 8153- PM 14.04 19.6 1.40 7883 PL + PB 14.31 29.4 2.05 Gamma 13.49 4886- PM 1.75 19.6 11.2 Shield Shell 5589 PL + PB 5.65 29.4 5.20 N11 (Hot) Inner Shell 13.85 8153- PM 13.39 19.6 1.46 Side Drop 7883 PL + PB 13.68 29.4 2.15 Gamma 13.44 4886- PM 1.73 19.6 11.33 Cylinder 5589 PL + PB 5.70 29.4 5.16 N12 (Hot) Inner Shell 13.75 224-254 PM 13.01 19.6 1.51 Vibration PL + PB 13.47 29.4 2.18 N13 Flange 13.39 4482- PM 4.89 19.6 4.01 (Cold) 4485 PL + PB 11.89 29.4 2.47 Vibration Inner Shell 14.64 373- PM 13.60 19.6 1.44 15591 PL + PB 14.02 29.4 2.10 Page 2-66

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-14 Linearized Stress Evaluation of Normal Condition of Transport Load Combinations (4 Sheets)

Nodal Linearized Stress Intensity Stress Factor Load Intensity Node Magnitude Allowable of Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety N14 Inner Shell 14.19 372- PM 13.42 19.6 1.46 (Hot) 15590 PL + PB 13.96 29.4 2.11 Shock N15 Flange 13.56 4482- PM 5.08 19.6 3.86 (Cold) 4485 PL + PB 12.04 29.4 2.44 Shock Inner Shell 15.08 372- PM 14.02 19.6 1.40 15590 PL + PB 14.57 29.4 2.02 Page 2-67

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-15 TN-40 Performance Evaluation Overview (Hypothetical Accident Conditions of Transport)

Loading Conditions SAR Section Scope of Evaluation 30 foot Free Drop 2.7.1.1 Cask body stresses due to bottom end drop 71.73-(c)(1) Cask body stresses due to lid end drop 2.7.1.2 Cask body stresses due to side drop 2.7.1.3 Cask body stresses due to CG over corner drop 2.7.1.4 Cask body stresses due to 20° slap down impact at lid end 2.7.1.5 Lid bolt analysis 2.7.1.6 Impact limiter attachment analysis Puncture 2.7.2 Cask body evaluation for 40 inch drop onto the 71.73-(c)(2) puncture bar Thermal 2.7.3.1 Maximum component pressures and temperatures 71.73-(c)(3) 2.7.3.2 Cask body thermal stresses due to fire accident 2.7.3.3 Maximum combined stresses Immersion 2.7.4.1 Cask body stresses due to 3 foot water head (1.3 psi) 71.73-(c)(5) 2.7.4.2 Cask body stresses due to 50 foot water head (21.7 71.73-(c)(6) psi) 71.61 2.7.4.3 Containment vessel stresses due to 290 psi external pressure (pressure directly applies to the containment vessel)

Buckling analysis of the containment vessel due to 290 psi external pressure Basket Evaluation 2.7.5 Structural analysis of the basket due to 30 foot end drop and 30 foot side drop loads Summary of Accident 2.7.6 Lists the highest stress intensities in the containment Condition Structural vessel and gamma shield shell and compare with the Analysis allowables Page 2-68

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-16 Summary of Individual Load Factors for Hypothetical Accident Condition of Transport Factor Used for Run Load Used in Individual Load Type Accident Conditions No. Analysis Load Combinations IL-1 Bolt preload and lid seating pressure - 1.0 IL-2 Fabrication Stresses - 1.0 IL-3 Internal pressure 100 psig 1.0 1.0 IL-4 External pressure 25 Thermal stresses at 100°F (hot)

IL-5 - 1.0 environment Thermal stresses at -20°F (cold)

IL-6 - 1.0 environment Cask Horizontal on Skid - 1g Down IL-8 1g 1.0 Gravity IL-12 End drop on bottom 1g 68 IL-13 End drop on lid 1g 68 IL-15 Side drop/slapdown 1g 75 CG Over Corner Drop on Front (Lid)

IL-16 32g axial,14g radial 1.28(1)

Impact Limiter CG Over Corner Drop on Rear (Bottom)

IL-17 32g axial, 14g radial 1.28(1)

Impact Limiter Notes:

(1) Taken from Section 2.7.1.2 Page 2-69

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-17 Summary of Load Combinations for Hypothetical Accident Condition of Transport Applicable Individual Load IL-1 IL-2 IL-3 IL-4 IL-12 IL-13 IL-15 IL-16 IL-17 Corner Load Bolt Internal External Bottom end Top end Side Corner drop drop Combination Combination Preload Fabrication Pressure Pressure Drop Drop drop Lid Bottom Number 30 Ft. End Drop on x x x x A1 Bottom End x x x x A2 30 Ft. End Drop on x x x x A3 Lid End x x x x A4 30 Ft. Side x x x x A5 Drop/Slapdown x x x x A6 30 Ft. CG Over x x x x A7 Corner Drop on Bottom End x x x x A8 30 Ft. CG Over x x x x A9 Corner Drop on Lid End x x x x A10 Fire Accident X X X A15 Page 2-70

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-18 Summary of Load Combination Stresses for Hypothetical Accident Condition of Transport Cask Component Nodal Stress Intensity (ksi)

Stress Inner Gamma Bottom Trunnion Allowable Load Comb. Type Lid Flange Shell Cylinder Plate Region (ksi)

A1 (Hot) 8.33 15.49 19.98 7.12 8.21 4.51 45.50 30 Ft. End Drop on Bottom End A2 8.90 16.27 20.85 6.80 6.17 4.15 45.50 (Cold)

A3 (Hot) 12.83 18.21 19.51 6.94 6.42 6.94 45.50 30 Ft. End Drop on Lid End A4 12.39 19.01 21.36 6.77 11.27 6.59 45.50 (Cold)

A5 (Hot) 39.92* 51.81* 18.93 62.29* 34.97* 15.27 45.50 30 Ft. Side Drop/Slapdown A6 39.91* 50.69* 19.01 62.28* 29.93 15.27 45.50 (Cold) 30 Ft. CG Over A7 (Hot) 8.45 14.48 33.21* 20.41 20.06 4.49 45.50 Corner Drop on A8 Bottom End 8.36 15.25 34.10* 20.38 20.00 4.14 45.50 (Cold) 30 Ft. CG Over A9 (Hot) 47.06* 37.80* 28.19 46.02* 15.83 8.59 45.50 Corner Drop on A10 Lid End 49.36* 40.56* 28.72 45.98*` 15.80 8.53 45.50 (Cold)

A15 Fire Accident 8.71 12.65 13.74 7.21 6.95 3.68 45.50 (600º F)

The stresses which result in a factor of safety less than 1.5 (stress higher than 45.5/1.5 = 30.3 ksi) are linearized in Table 2-19.

Page 2-71

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-19 Linearized Stress Evaluation for Hypothetical Accident Condition Load Combinations (2 Sheets)

Nodal Stress Linearized Stress Intensity Intensity Node Magnitude Allowable Factor of Load Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety Lid & Shield 39.92 15933- PM 13.57 45.50 3.35 15927 PL + P B 27.73 65.00 2.34 A5 (Hot) Flange 51.81 15719- PM 21.91 45.50 2.08 (30 Ft. Side 458 Drop/Slapdown PL + P B 36.84 65.00 1.76 Gamma 62.29* 4886- PM 7.66 45.50 5.94 Shield Shell 5589 PL + P B 24.61 65.00 2.64 Lid & Shield 39.91 15933- PM 13.54 45.50 3.36 15927 PL + P B 27.70 65.00 2.35 A6 (Cold) Flange 50.69 15719- PM 21.95 45.50 2.07 30 Ft. Side 458 Drop/Slapdown PL + P B 36.64 65.00 1.77 Gamma 62.28* 4793- PM 7.67 45.50 5.93 Shield Shell 5649 PL + P B 24.60 65.00 2.64

The nodal stress intensity at this location is highly local and results in low linearized membrane and bending stresses at the cross section. However, there is high peak stress at this location.

Page 2-72

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2-19 Linearized Stress Evaluation for Hypothetical Accident Condition Load Combinations (2 Sheets)

Linearized Stress Intensity Nodal Stress Node Magnitude Allowable Factor of Load Comb. Component Intensity (ksi) Nos. Type (ksi) (ksi) Safety A7 (Hot) PM 12.02 45.50 5.69 30 Ft. CG over Inner Shell 33.21 202-246 Corner Drop on PL + P B 17.99 65.00 3.63 Bottom End A8 (Hot) PM 11.94 45.50 5.89 30 Ft. CG over Inner Shell 34.10 202-246 Corner Drop on PL + P B 18.38 65.00 3.57 Bottom End Lid & Shield 4537- PM 12.42 45.50 3.66 47.06 4548 PL + P B 34.77 65.00 1.87 A9 (Hot) 30 Ft. CG over Flange 3922- PM 8.90 45.50 5.11 37.80 Corner Drop on 5429 PL + P B 21.67 65.00 3.00 Lid End Gamma Shield PM 6.23 45.50 7.30 4793-Shell 46.02 5649 PL + P B 18.01 65.00 3.61 Lid & Shield 4537- PM 12.46 45.50 3.65 49.36 4548 PL + P B 35.36 65.00 1.84 A10 (Cold) 30 Ft. CG over Flange 3922- PM 9.19 45.50 4.95 40.56 Corner Drop on 5429 PL + P B 22.89 65.00 2.84 Lid End Gamma Shield PM 6.27 45.50 7.26 4793-Shell 45.98 5649 PL + P B 18.02 65.00 3.61 Page 2-73

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2-1 Geometry of Upper (front) and Lower (rear) Trunnions Page 2-74

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2-2 290 psig Immersion Analysis Finite Element Model Loads and Boundary Conditions Page 2-75

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2-3 DELETED Figure 2-4 DELETED Figure 2-5 DELETED Figure 2-6 DELETED Page 2-76

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.1 STRUCTURAL ANALYSIS OF CASK BODY TABLE OF CONTENTS 2.10.1.1 Introduction........................................................................ 2.10.1-1 2.10.1.2 ANSYS Analysis ................................................................ 2.10.1-2 2.10.1.3 ANSYS Analysis Results and Report Methodology ......... 2.10.1-22 2.10.1.4 Trunnion Local Stress Analysis Due To Lifting Load ....... 2.10.1-23 2.10.1.5 References ...................................................................... 2.10.1-24 LIST OF TABLES Table 2.10.1-1 Element types, Material Numbers and Real Constants of ANSYS Model Cask Components ........................................ 2.10.1-25 Table 2.10.1-2 Summary Maximum Nodal Stress Intensities in Cask Components for Individual Load Runs .................................. 2.10.1-26 Table 2.10.1-3 6g Lifting Load ...................................................................... 2.10.1-27 Page 2.10.1-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 2.10.1-1 TN-40 Containment Vessel Key Dimensions ........................ 2.10.1-28 Figure 2.10.1-2 TN-40 Cask/Lid/Bolt FEM Representation ............................ 2.10.1-29 Figure 2.10.1-3 Finite Element ModelLid .................................................... 2.10.1-30 Figure 2.10.1-4 Finite Element Model - Containment Flange ........................ 2.10.1-31 Figure 2.10.1-5 Finite Element Model - Inner Shell and Bottom Plate ........... 2.10.1-32 Figure 2.10.1-6 Finite Element Model - Gamma Shield Shell........................ 2.10.1-33 Figure 2.10.1-7 Finite Element Model - Gamma Bottom Shield .................... 2.10.1-34 Figure 2.10.1-8 Coupling and Boundary Conditions ...................................... 2.10.1-35 Figure 2.10.1-9 Bolt Preload - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-36 Figure 2.10.1-10 Fabrication Load - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-37 Figure 2.10.1-11 Internal Pressure - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-38 Figure 2.10.1-12 Thermal Stress 100 °F Environment - Loading and Displacement Boundary Conditions ...................................... 2.10.1-39 Figure 2.10.1-13 Thermal Stress -20 °F Environment - Loading and Displacement Boundary Conditions ...................................... 2.10.1-40 Figure 2.10.1-14 Thermal Stress -40 °F Environment - Loading and Displacement Boundary Conditions ...................................... 2.10.1-41 Figure 2.10.1-15 1g Horizontal - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-42 Figure 2.10.1-16 Transport Tie-Down - Loading and Displacement Boundary Conditions ............................................................ 2.10.1-43 Figure 2.10.1-17 End Drop on Bottom - Loading and Displacement Boundary Conditions ............................................................ 2.10.1-44 Figure 2.10.1-18 End Drop on Lid - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-45 Figure 2.10.1-19 6g Vertical Lifting - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-46 Figure 2.10.1-20 Side Drop, 1g - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-47 Figure 2.10.1-21 Corner Drop on Lid - Loading and Displacement Boundary Conditions ............................................................................. 2.10.1-48 Figure 2.10.1-22 Corner Drop on Bottom - Loading and Displacement Boundary Conditions ............................................................ 2.10.1-49 Page 2.10.1-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-23 20° Slap Down Impact On Lid End - Loading and Displacement Boundary Conditions ...................................... 2.10.1-50 Figure 2.10.1-24 20° Slap Down Impact on Bottom End- Loading and Displacement Boundary Conditions ...................................... 2.10.1-51 Page 2.10.1-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 STRUCTURAL ANALYSIS OF CASK BODY 2.10.1.1 Introduction This appendix presents the structural analyses of the TN-40 cask body and the local stress at the trunnion/cask body interface. The cask body includes the inner shell, bottom inner plate, gamma shield shell, bottom shield, shield plate, and the lid outer plate. The methods, models, and assumptions used in analyzing the cask body under various individual loading conditions as specified in 10CFR71.71 and 10CFR71.73 [1]

are described. These conditions include both the Normal Conditions of Transport (NCT) and the Hypothetical Accident Conditions (HAC). Stress results are reported at selected locations for each load case. Maximum stresses from this appendix are evaluated in Sections 2.6 and 2.7 where the load combinations as outlined in Regulatory Guide 7.8 [3] are performed and the results are evaluated against the ASME Code [4] and Regulatory Guide 7.6 [5] design criteria, which are described in Section 2.1.2.

Static linear elastic methods are used for the TN-40 cask body structural analyses. The stresses and deformations resulting from the applied loads are generally determined using the ANSYS [6] computer program.

The detailed calculations for the lid bolts are presented in Appendix 2.10.2. Stress evaluations of the lifting devices and tie-down system are described in Section 2.5.

The two analysis methods described in this appendix and used to evaluate the cask body for the specified loading conditions are:

ANSYS Analysis - Static Linear Elastic Analysis using a 3-D Model

Bijlaard Trunnion Local Stress Analysis The Bijlaard [7] analyses are performed to determine the local cask body stresses at trunnion locations where general stresses are also reported from the ANSYS analyses.

This permits the localized shell stresses induced by the trunnion loadings to be easily combined with stresses obtained from appropriate ANSYS load cases. The method of combining stress results from individual load cases and their evaluations are discussed in Section 2.6 and Section 2.7 for NCT and HAC loads, respectively.

Page 2.10.1-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.1.2 ANSYS Analysis Cask Geometry Description The cask body consists of an inner shell and an outer gamma shield shell. The inner shell (and bottom inner plate) is the primary containment boundary of the packaging.

Key dimensions of the cask body are shown in Figure 2.10.1-1. The inner shell is 1.5 in. thick cylinder welded to the shell flange and bottom inner plate. The inner shell is shrunk fit into the gamma shield shell. The cask lid is bolted to the shell flange by 48, 1.5 in. diameter, high strength bolts and sealed with two metallic seals. The lid, inner shell, bottom inner plate, and gamma shield shell components are made of low alloy steel forgings and plates. The cask is fitted with an impact limiter at either end. The two impact limiters are held against the cask ends by a set of tie rods connecting them.

Two sets of trunnions are welded to the side of the gamma shield shell upper and lower ends for handling and supporting the cask during lifting and handling operations. A basket assembly inside the cask cavity is used to position and support the fuel assemblies. A detailed physical description of the containment components is provided in Chapter 1. Appendix 1.4 contains reference drawings of the TN-40 package on which the analysis models are based.

ANSYS Cask Model The gamma shield shell, the inner shell and bottom inner plate, the shell flange, and the lid and its shield are modeled utilizing ANSYS eight-node brick elements (SOLID45) as shown in Figure 2.10.1-2 to Figure 2.10.1-7. Due to the cyclic symmetry of the TN-40 body, some nodes in the FEM are rotated into a local cylindrical coordinate system for easy application of node coupling and boundary conditions. This local coordinate system is located at the model axis of symmetry. The lid bolts are modeled using BEAM4 elements. The bolt preload is simulated using pre-strain in the beam elements by their real constants. The nodes at common surfaces between flange and lid are coupled in the model axial direction. The element nodes on the shrunk-fit surfaces between the gamma shield shell and the inner shell are coupled in the radial direction and are shown in Figure 2.10.1-8. Similarly, the nodes on the contacting surface between the bottom shield and the bottom inner plate are coupled in the axial direction.

A total of 18,688 elements and 24,354 nodes comprise the ANSYS finite element model of the TN-40 cask. At the cask model cut plane, symmetry boundary conditions are applied. The element types, material numbers and real constants of different components of the model are summarized in Table 2.10.1-1. For various loading conditions, different boundary conditions are used to avoid rigid body motion of the FEM. The loadings and displacement boundary conditions for individual load cases are shown in Figure 2.10.1-9 to Figure 2.10.1-24. To aid clarity, symmetry boundary conditions (UZ) are not shown in these figures.

Page 2.10.1-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Material Properties The materials used for TN-40 transport cask and their properties as a function of temperature are listed below [4].

Youngs Ultimate Yield Allowable Thermal Cask Temperature Modulus Material Strength Strength Sm Expansion Component (°F) E (ksi) (ksi) (ksi) (in/in/°F)

(psi) 70 65 37 21.7 27.8x106 6.27x10-6 Containment Inner Shell & SA-203 Gr.

200 65 33.9 19.6 27.1x106 6.54x10-6 Bottom Inner E or Gr. D Plate 300 65 32.7 19.6 26.7x106 6.78x10-6 70 70 37.5 23.3 27.8x106 6.27x10-6 SA350 Flange and LF3 or SA- 200 70 34.3 22.8 27.1x106 6.54x10-6 Lid Outer Plate 203 Gr. E 300 70 33.2 22.2 26.7x106 6.78x10-6 70 70 36 23.3 29.5x106 5.73x10-6 SA105 or Lid Shield SA516, 200 70 33 21.9 28.8x106 6.09x10-6 Gr.70 300 70 31.8 21.3 28.3x106 6.43x10-6 SA266 CL 70 70 36 23.3 29.5x106 5.73x10-6 Gamma Shield 4 or Shell Cylinder & SA516, 200 70 33 21.9 28.8x106 6.09x10-6 Bottom Shield Gr.70 or SA-105 300 70 31.8 21.3 28.3x106 6.43x10-6 70 125 105 35 27.8x106 6.27x10-6 SA320 Gr.

Lid Closure Bolt 200 125 99 33 27.1x106 6.54x10-6 L43*

300 125 95.7 31.9 26.7x106 6.78x10-6 Note: Lower strengths of alternate materials are listed in the above table.

The lid bolt material specified in SAR drawing 10421-71-1 is SA540 Gr. B23/B24 CL1. The only SA320 Gr. L43 material property used in this appendix is Youngs modulus which is identical to that of SA540 Gr. B23/B24 CL1.

Thus the results provided remain valid. The lid bolt evaluation concerning delayed impact presented in Appendix 2.10.11 is based on SA540 Gr. B23/B24 CL1.

Page 2.10.1-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Loadings This analysis evaluates NCT and HAC loadings as specified in 10CFR71. The 20 individual load cases considered in this evaluation are described below.

TN-40 24 Individual Load Cases IL-1 Bolt Preload and Lid Seating Pressure IL-2 Fabrication Stress IL-3 Internal Pressure (100 psig)

IL-4 External Pressure (25 psig)

IL-5 Thermal Stress Due to Hot Environment (100°F)

IL-6 Thermal Stress Due to Cold Environment (-20°F)

IL-7 Thermal Stress Due to Cold Environment (-40°F)

IL-8 Horizontal Cask Supported by Skid, 1g Down Gravity Load IL-9 Transport Tie down Load (10g Long., 5g Lat., 2g Vert.)

IL-10 Transport Rail Vibration Load Supported by Skid (0.19g Long., 0.19g Lat., 0.37g Vert.)

IL-11 Transport Rail Shock Load Supported by Skid (4.7g All Directions)

IL-12 End Drop on Bottom -Rear (Bottom) Impact Limiter (1g)

IL-13 End Drop on Lid - Front (Top) Impact Limiter (1g)

IL-14 6G on Front Trunnion Lifting Load (Cask Vertical, 6g Up)

IL-15 Side Drop (1g)

IL-16 CG Over Corner Drop on Lid End (32g Axial, 14g Radial)

IL-17 CG Over Corner Drop on Bottom End (32g Axial, 14g Radial)

IL-18 20° Slap Down Impact on Lid End (22g Axial, 39g Radial)

IL-19 20° Slap Down Impact on Bottom End (22g Axial, 39g Radial)

IL-20 Local Stresses at Upper Trunnion/Cask Body Interface with 1g up - Cask Vertical The magnitudes of the loads and pressures used in each individual load case analysis are computed as described in the following paragraphs based on the following TN-40 weights:

Calculated Component Weight Used (lbs.)

Weight (lbs.)

Cask Body 169,796 169,800 Internals 66,693 66,700 Top (Front) Impact Limiter and Top 17,489 17,700 Impact Limiter Spacer Bottom (Rear) Impact Limiter 16,332 16,500 Tie Rods and Bolting Brackets 1,145 1,000*

Total 271,455 271,700

This weight is equally divided in two impact limiters in the analysis. i.e.

Front Impact limiter Weight = 17,700 + 500 = 18,200 LB Rear Impact limiter Weight = 16,500 + 500 = 17,000 LB Page 2.10.1-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

1. Lid Closure Bolt Preload and Lid Seating Pressure (IL-1)

A bolt axial prestress of 50 ksi, calculated in Appendix 2.10.2, at the bolt shank (1.375 in. diameter) is simulated by specifying an initial strain in BEAM4 elements representing the bolts. The required initial strain value of 0.002108 in./in. (in the bolts) was determined by first calculating the initial strain required to produce an axial stress of 50 ksi (i.e. =/E = 50E3/26.7E6 = 0.001873 in./in.). Then, an initial analysis with the calculated strain (0.001873) was conducted and the resulting bolt prestress was backed out. Since, a portion of this strain becomes elastic preload strain in the bolts, and a portion becomes strain in the clamped parts, the backed out prestress from the initial analysis will not produce the desired 50 ksi. The FEM was then updated by multiplying the 0.001873 in./in. strain by the ratio of (desired prestress, 50 ksi / initial analysis prestress), and a second preload analysis was conducted, which resulted in a 50 ksi bolt prestress.

The maximum lid seating pressure required to seat the metallic seals [8] is computed as 1,038,108 lb. This load is calculated based on approximately 2200 lb/in. gasket seating force. Based on the closest nodes in the model to the seal location, the lid seating pressure, p = 1,038,108 / (38.42 - 37.352) = 4154.5 psi This pressure is applied to the lid and flange seal areas. During the analysis, the cask is supported as shown in Figure 2.10.1-9.

2. Fabrication Stress (IL-2)

The fabrication stresses in the cask are due to the 0.040 in. diametrical interference between the inner shell and the gamma shield shell (Appendix 1.4 drawings). The shrink fit stresses are calculated based on a radial interference of 0.020 in. The interface pressure, p, can be found from the following expression:

p = (E/b) [(b2-a2) (c2-b2)/2b2(c2-a2)] [10]

where:

= 0.20 in. E = 29.5x106 psi a = 36.0 in. b = 37.5 in. c = 45.5 in.

Page 2.10.1-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 a b c Inner Shell Gamma Shield Shell p = (29.5x106 x .02/37.5) [(37.52-36.02) (45.52-37.52)/2x37.52(45.52-36.02)]

= 529.1 psi Use 530 psi.

In order to model this load, the radial couplings between the inner shell and gamma shield shell are deleted and the interface 530 psi pressure is applied to the two cylinders. The boundary conditions for this loading are shown in Figure 2.10.1-10.

3. Internal Pressure Loading (IL-3)

An internal pressure of 100 psig is applied to the cavity surface as shown in Figure 2.10.1-11. The pressure is applied up to the metallic seal inner radius. The lid closure bolt preload and seal seating loads are removed in this calculation. The cask is supported as shown in Figure 2.10.1-11 for this loading.

4. External Pressure Loading (IL-4)

An external pressure of 25 psig is applied to the outer surface of the cask body. The pressure is applied up to the seal outer radius.

5. Thermal Stress for Hot Environment Condition at 100F Ambient Temperature (IL-5)

The thermal analysis of the cask body is described in Chapter 3. The thermal model is used to obtain the NCT steady state component temperatures in the cask body.

The thermal inputs include 100F daily averaged ambient air temperature, maximum payload decay heat and maximum solar heat loading. The cask nodal temperatures from the thermal results file are interpolated using an ANSYS macro to determine the nodal temperatures in the structural model. The resulting cask temperature distribution is shown in Figure 2.10.1-12. These temperatures are then used as ANSYS inputs for the thermal stress analysis. Temperature-dependent material properties are used in this analysis. The cask support boundary conditions are also shown in Figure 2.10.1-12.

Page 2.10.1-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

6. Thermal Stresses for Cold Environment Condition at -20F Ambient Temperature (IL-6)

The thermal analysis of the cask body is described in Chapter 3. The thermal model is used to obtain the NCT steady-state temperatures in the cask body due to a -20F daily averaged ambient air temperature. The cask nodal temperatures from the thermal results file are interpolated using an ANSYS macro to determine the nodal temperatures in the structural model. The resulting cask temperature distribution is shown in Figure 2.10.1-13. These temperatures are then used as ANSYS inputs for the thermal stress analysis. Temperature-dependent material properties are used in this analysis. The cask support boundary conditions are also shown in Figure 2.10.1-13.

7. Thermal Stresses for Cold Environment Condition at -40F Ambient Temperature (IL-7)

The thermal analysis of the cask body is described in Chapter 3. The thermal model is used to obtain the NCT steady-state metal temperatures in the cask body resulting from the -40F daily averaged ambient air temperature. The cask nodal temperatures from the thermal results file are interpolated using an ANSYS macro to determine the nodal temperatures in the structural model. The resulting cask temperature distribution is shown in Figure 2.10.1-14. These temperatures are then used as ANSYS inputs for the thermal stress analysis. Temperature-dependent material properties are used in this analysis. The cask support boundary conditions are also shown in Figure 2.10.1-14.

8. Cask Supported Horizontally by Skid, 1g down Gravity Load (IL-8)

For the 1g loading, the cask is oriented horizontally, and the cask is secured axially and radially on a transport skid. For the inertial loading, a vertical acceleration of 1g is applied in the global X direction.

The weight of the internals is applied as radial pressure (Pr), the cosine varying pressure is applied around the lower radial portion (0° to 75° range) of the cavity.

The pressure is calculated by the following formula, which is developed in Section 2.10.1.2.1:

Where:

W 1 i = 1/2 Angle of contact Pr = [g] [ ] cos( ) i = Circumferential angle pressure LR sin( + ) sin( ) 2 is applied 2 + 2 W = Weight of internals L = Length pressure is applied

( ) +1 ( ) 1 2 2 R = Radius pressure is applied g = Vertical acceleration Page 2.10.1-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 For example, a pressure applied at an angle of 7.5° would be calculated as follows:

66,700 1 180 7.5 Pr = [1.0] [ ] cos( )

163.0 36 sin(90 + 75) + sin(90 75) 2 75 180 180

( ) +1 ( ) 1 2 75 2 75 Pr = 11.367(0.7083) (0.9877) = 7.95 psi In addition, radial pressure due to the front impact limiter weight (Pfr) is applied along the contacting surfaces of the limiter and the lid/cask wall. This includes the lid and flange radial surfaces. The pressure follows a cosine distribution and is applied from the vertical (Y=180º to Y=105º).

A radial pressure due to the rear impact limiter weight (Prr) is also applied along the contacting surfaces of the limiter and the cask wall. Similar to the front impact limiter, the pressure follows a cosine distribution and is applied from the vertical (Y=180º to Y=105º).

Since, resin (10,577 lb.), outer shell (7,453 lb.), trunnions (666 lb.), and aluminum box (1,993 lb.) weights are not included in the FEM, their weight is incorporated in outer gamma cylinder (mat, 6) by increasing its actual density (0.283) to an equivalent density of 0.343 lb/in3. This increased density is used in all subsequent drop and inertia load runs.

During the run, the cask is supported as shown in Figure 2.10.1-15. Since the support saddle and strap are not modeled, displacement boundary conditions are applied to nodes near these locations.

9. Transport Tie-Down Loading (IL-9)

For the tie down loading, the cask is oriented horizontally and secured axially and radially on a transport skid. The input loading conditions used to evaluate the TN-40 cask for transport tie-down loading are obtained from 10CFR71.45. The peak inertia (acceleration) values used are:

Vertical 2g Longitudinal 10 g Lateral 5g Two inertial loads are applied in the FEM:

1. A longitudinal 10g acceleration (applied in the axial direction)

2. The resultant of the vertical & lateral accelerations (applied in the radial direction) is calculated as (22 + 52)1/2 = 5.4g.

A pressure due to the weight of the front impact limiter (Pfa) is applied axially at lid end and calculated as:

Pfa = 10.0 x (18,200) / ((3.14159)*45.52) = 27.98 psi Page 2.10.1-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 In addition, pressure loads due to the weight of the internals are applied to the cask inner surface in the axial and radial directions. The pressure due to the axial load acting on the inside surface of the rear bottom plate (Pia) is calculated by:

Pia = 10.0 x (66,700) / ((3.14159)*36.02) = 163.82 psi Radial pressure (Pr) acting on the lower half of the inner cask surface due to the weight of internals is represented as a cosine varying pressure around the lower radial portion (0° to 75° range) of the cavity as described above.

As an example, pressure applied at an angle of 7.5° would be calculated as follows:

66,700 1 180 7.5 Pr = [5.4] [ ] cos( )

163.0 36.0 sin(90 + 75) sin(90 75) 2 75

+

180 180

( ) +1 ( ) 1 2 75 2 75 Pr = 61.380 * (0.7083)(0.9877) = 42.94 psi In addition, the radial pressure due to the front impact limiter weight (Pfr) is applied along the contacting surfaces of the limiter and the lid/cask wall. This includes the lid and flange radial surfaces. The pressure follows a cosine variation and is applied from the vertical (Y=180º to Y=105º).

A radial pressure due to the rear impact limiter weight (Prr) is also applied along the contacting surfaces of the limiter and the cask wall. Similar to the front impact limiter, the cosine varying pressure is applied from the vertical (Y=180º to Y=105º).

During the ANSYS analysis, the cask is supported as shown in Figure 2.10.1-16.

Since the skid saddles are not modeled, displacement boundary conditions are applied to nodes near the saddle centerline locations. Axial restraint is provided by displacement boundary conditions at the cask bottom (rear) nodes.

10 & 11 Rail Car Vibration and Rail Car Shock Loadings (IL-10 & IL-11)

For the rail car vibration and shock loadings, the same methodology utilized for the transport tie-down loading is applied, with the exception that the inertial loads are based on the following accelerations:

Rail Car Vibration Accelerations [9]:

Vertical 0.37g Longitudinal 0.19g Lateral 0.19g Rail Car Shock Accelerations [9]:

Vertical 4.7g Longitudinal 4.7g Lateral 4.7g Page 2.10.1-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

12. End Drop on Bottom (Rear) Impact Limiter (IL-12)

The dynamic analysis described in Appendix 2.10.8 determines the inertial load on the TN-40 packaging for both a 1 foot and a 30 foot end drop onto an unyielding surface. This stress evaluation is conducted for a unit load (1g). However, since this is a linear elastic analysis, stresses can be ratioed for the actual g loads when the load combinations are calculated. The payload and the impact limiters are not included in the FEM. Rather, their loading effects are simulated as distributed pressures applied to the cask at the appropriate locations.

The following inertia loads (pressure) are applied due to a 1g vertical acceleration.

An axial pressure due to the cask internals (Pi) is applied to the bottom inner plate surface:

Pi = 1.0 x 66,700/(() x 36.02) = 16.382 psi An axial pressure due to the front impact limiter (Pfl) is applied to the outer lid surfaces based on the projected area:

Pfl = 1.0 x 18,200/(() x 45.52) = 2.798 psi The bottom nodes of the cask are supported in vertical direction. Loading and displacement boundary conditions are shown in Figure 2.10.1-17.

13. End Drop on Lid (Front) Impact Limiter (IL-13)

An analysis similar to that of bottom end drop is performed for the 1g vertical load.

The following loads are applied.

A 1.0 g vertical acceleration of the finite element model simulates the inertial loading.

An axial pressure due to the cask internals (Pi) is applied to the inner lid surface based on:

Pi =1.0 x 66,700/(() x 36.02) = 16.382 psi An axial pressure due to the rear impact limiter (Prl) is applied to the outer surface of the cask bottom shield:

Prl = 1.0 x 17,000/(() x 45.52) = 2.614 psi Loading and displacement boundary conditions are shown in Figure 2.10.1-18.

14. 6g Lifting on Upper Trunnion (IL-14)

The cask is oriented vertically and supported by the 2 upper trunnions. The inertial loading is simulated by applying a 6g vertical acceleration to the finite element model. Note that the impact limiters are not included in this case, as they are removed prior to lifting.

Page 2.10.1-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Since the internals are not included in the model, their loading effects are simulated by a distributed pressure (Pi) acting on the inside bottom surface of the cask cavity:

Pi = 6.0 x (66,700) / ((3.14159)*36.02) = 98.293 psi Loading and displacement boundary conditions are shown in Figure 2.10.1-19. This load step is to calculate the globe stress of the cask component and will be used for containment fatigue analysis. The stresses at the trunnion/gamma shield shell interface due to 6g lifting load are calculated in Section 2.10.1.4.

15. Side Drop (IL-15)

The dynamic analyses described in Appendix 2.10.8 determine the inertial loads on the TN- 40 packaging for both 1 foot and 30 foot side drops onto an unyielding surface. This stress evaluation assumes a unit load (1g). However, since this is a linear elastic analysis, stresses can be ratioed for the actual g loads when the load combinations are calculated. The payload and the impact limiters are not included in the FEM. Their loading effects are simulated as distributed pressures applied to the cask at the appropriate locations.

The contacting impact limiter forces on the cask and lid are applied as reaction pressures required to balance the inertial forces of the system. Thus, the vessel is in equilibrium under the applied forces. During the side drop, the pressure on the inner surface due to the internals and the reaction pressure on the outer cask surface due to the impact limiters are assumed to vary as a cosine function over a defined arc length.

The loads acting in this case are:

A. Cask Body Inertia The inertial loading is simulated by applying a 1g vertical acceleration to the finite element model in the global X direction which is perpendicular to the cask axis.

B. Pressure Due to Internals The radial pressure (Pi) acting on the lower half of the inner shell surface due to the weight of internals is represented as a cosine varying pressure applied around the lower radial portion (0° to 75° range) of the cavity surface.

W 1 i Pi = [g] [ ] cos( )

LR 2 sin( + ) sin( )

2 + 2

( ) +1 ( ) 1 2 2 For example, a pressure applied at an angle of 7.5º would be calculated as follows:

Page 2.10.1-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 66,700 1 180 7.5 Pi = [1.0] [ ] cos( )

163.0 36.0 sin(90 + 75) sin(90 75) 2 75

+

180 180

( ) +1 ( ) 1 2 75 2 75 Pi = 7.95 psi C. Impact Reaction Pressures:

Pressures applied by the rear and front impact limiter reactions on the lower longitudinal half of the outer cask body during impact are computed. These radial pressures are assumed to vary in a cosine distribution around the bottom half of the outer surfaces (0° to 89.5° range) and are calculated just as the internals pressure above. However, the total force (F) applied in the equation is based on the following reactions:

Total cask weight, W = 236,500 lb. Use 237,000 lb. (cask + internals)

Reaction force, Front (lid) = 237,000 x 92.33/183.75

= 118,974 lb.

Reaction force, Rear (bottom) = 237,000 x 91.42/183.75

= 118,026 lb.

The front (lid / cask side) reaction force is divided in the ratio of two lengths:

4.5 in. (R = 41.375) for the lid and 7.5 in. (R = 45.5) for the portion of the cask covered by the impact limiter.

Loading and displacement boundary conditions are shown in Figure 2.10.1-20.

16. 30 Foot CG Over Corner Drop on Lid End (IL-16)

For CG over corner, the cask is inclined at approximately 64º from the horizontal.

The dynamic analysis of Appendix 2.10.8 determines the inertial loads for this loading condition. All the applied loads and reaction forces are transformed into axial and normal components. The axial pressure components due to the internals, bottom impact limiter and impact reaction are assumed uniformly distributed. All radial pressure components (i.e. pressure due to internals, rear impact limiter and impact reactions) are assumed to have a cosine variation over a defined arc length.

Page 2.10.1-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The forces acting in this case are:

A. Cask Body Inertia The component accelerations (32g axial & 14g Radial - Appendix 2.10.8) are applied as inertial loads in the axial and radial directions. In addition, a rotational acceleration of 0.148g is applied at the vessel CG to counteract the out-of-balance caused by the components acceleration resultant. That is, the component translational accelerations applied have been conservatively rounded, which results in a slight resultant moment (out-of-balance) when the solution is executed. This moment is counteracted by the applied angular acceleration (torque) to preserve static equilibrium.

B. Pressure Due to Internals Radial pressure (Pir) acting on the lower half of the inner cask wall due to the weight of internals is represented as a cosine varying pressure around the upper radial portion (180° to 105° range) of the cavity.

In addition, an axial pressure is applied, due to the weight of the internals, to the cask inner lid surface:

Pia = 32 x (66,700) / ((3.14159)*36.02) = 524.2 psi C. Pressure Due to Rear Impact Limiter The inertia load of the nonstriking impact limiter is also applied to the cask in two mutually perpendicular directions. The axial component (Pra) is applied as a uniform pressure over the outside surface at the interface with the impact limiter on the bottom end. The pressure applied is calculated as:

Pra = 32.0 x (17,000) / ((3.14159)*45.52) = 83.7 psi The other component (Prr) follows a cosine distribution around the lower half of the outside surface (0° to 75° range) of the cask.

D. Reaction Pressures Due to Front Impact Limiter The reaction pressure from the striking impact limiter is applied to the cask in two mutually perpendicular directions. The axial component (Pfa) is applied as two-step uniform pressure over a portion of the cask interface with the impact limiter on the lid end. The crush footprint of the front impact limiter was projected onto the cask surface based on results obtained from Appendix 2.10.8.

Page 2.10.1-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 45.5 A1 A2 A3 20 54.18 For the axial reaction pressure applied, a total area (4037 in2), for a 360° arc, is calculated as:

A = (() x 45.52)-A3 A3 = 1/2(R2) [2 -Sin2] [2]

Cos = 8.68/45.5 = 0.1908 = 79.0 deg = 1.3789 rad.

A3 = 1/2(45.52) [2x 1.3789 -Sin (2x79)] = 2466.9 in2 Total crushing area, A = (() x 45.52) - 2466.9 = 4037 in2 Due to the nature of a corner impact the reaction pressure will not be precisely distributed uniformly throughout the crushed area. Instead, the reaction pressures at the center of the crushed area will be higher than those in the peripheral area away from the impact center. It is therefore assumed that the majority weight consisting of the cask plus the rear end impact limiter is to be reacted by an area bounded by 20 in. from the edge of the cask and the weight of internal cargo is to be reacted by the rest of the crushed area. The axial reaction pressure to the cask is therefore applied in two steps: p1 on area A1 and p2 on area A2 the following formulas are used to calculate p1 and p2:

Cos = 25.5/45.5 = 0.5604 = 55.91 deg = 0.9759 rad.

A1 = 1/2(45.5 ) [2x 0.9759 -Sin (2x55.91)] = 1059.3 in2 2

A2 = 4037 -1059.3 = 2977.3 in2 Pressure on area A1:

p1 = 32.0 x (169,800+17,000) / 1059.3 in2 = 5,643 psi Pressure on area A2:

p2 = 32.0 x (66,700) / 2977.3 in2 = 718 psi Page 2.10.1-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The radial component pressures follow a cosine distribution around the radial crush foot print from 90.5o to 180o of the cask. The radial reaction pressures are calculated using a modified version of cosine formula. Since the crush footprint is a circular segment, and the pressures are being applied to two separate side surfaces (i.e. lid and upper flange wall). The total force (F) applied in the equation is based on the percentage of the total length to which the specific pressure is applied.

L F 1 (180 - i)

Pfr1 = [g] [ ] [ R ] [ ] cos( )

Lt LR 2 sin( + ) sin( )

2 + 2

( ) +1 ( ) 1 2 2 Page 2.10.1-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The calculated reaction axial and radial pressures due to the crushed impact limiter had to be adjusted and a rotational acceleration 0.148g applied to balance the applied loads. Loading and displacement boundary conditions are shown in Figure 2.10.1-21.

17. 30 Foot CG Over the Corner Drop on Bottom End (IL-17)

For this corner drop, the cask is again inclined at approximately 64º from the horizontal as described in Appendix 2.10.8. The applied loads are transformed into axial and normal components, and are applied using the same methodology adopted for the CG over corner lid drop. All radial pressure components (i.e. pressure due to internals, front impact limiter and impact reactions) are assumed to have a cosine variation. The forces acting in this case are:

A. Cask Body Inertia The component accelerations (i.e. 32g axial & 14g Radial - Appendix 2.10.8) are applied as translational inertial loads in the axial and radial directions respectively. In addition, a rotational acceleration of 0.1452g is applied at the vessel CG to counter act the out-of-balance and return the model to static equilibrium.

B. Pressure Due to Internals The radial pressure (Pir) acting on the lower half of the inner cask wall due to the weight of the cask internals is represented as a cosine varying pressure around the lower radial portion (0o to 75o range) of the cavity.

In addition, an axial pressure due to the weight of the internals is applied to the cask inner bottom surface based on:

Pia = 32.0 x (66,700) / ((3.14159)*36.02) = 524.2 psi C. Pressure Due to the Front Impact Limiter The inertia load of the nonstriking impact limiter is also applied to the cask in two mutually perpendicular directions. The axial component (Pla) is applied as a uniform pressure over the outer surface of the lid. The pressure applied is calculated as:

Pla = 32.0 x (18,200) / ((3.14159)*45.52) = 89.5 psi The other component (Plr) is assumed to follow a cosine distribution around the upper half of the outside surface (105o to 180o range) of the cask.

D. Reaction Pressures Due to Rear Impact Limiter The reaction pressure from the striking impact limiter is applied to the cask in two mutually perpendicular directions as described in the above Loading Case IL-16.

Page 2.10.1-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 For the axial reaction pressure applied, the total area, 4,037 in2, and areas A1 and A2 are the same as calculated for the corner drop on the lid (Case IL-16).

Pressure on area A1:

p1 = 32.0 x (169,800+18,200) / 1059.3 = 5,679 psi Pressure on area A2:

p2 = 32.0 x (66,700) / 2977.3 = 718 psi The radial component pressure (Prr) is assumed to follow a cosine distribution around the radial crush foot print from 0o to 89.5o of the cask. The radial reaction pressures are calculated using a modified version of the cosine distribution, based on the calculated angle of application.

The calculated reaction axial and radial pressures due to crushed impact limiter had to be adjusted and a rotational acceleration 0.1452g applied to balance the applied loads. Loading and displacement boundary conditions are shown in Figure 2.10.1-22.

18. 20° Slapdown Impact on Lid End (IL-18)

For the oblique lid impact, the cask is inclined at approximately 20º from the horizontal as described in Appendix 2.10.8. All the applied loads and reaction forces are transformed into axial and normal components respectively. The axial pressure components due to the internals, bottom impact limiter and impact reaction are assumed to be uniformly distributed in two steps. All radial pressure components (i.e. pressure due to internals, rear impact limiter and impact reactions) are assumed to have a cosine variation over a determined arc length.

The forces acting in this case are:

A. Cask Body Inertia The component accelerations (i.e. 22g axial & 39g Radial - Appendix 2.10.8) are applied as inertial loads in the axial and radial directions respectively. In addition, a rotational acceleration of 1.113g is applied at the vessel C.G. to counter act the out-of-balance forces and return the model to static equilibrium.

B. Pressure Due to Internals Radial pressure (Pir) acting on the lower half of the inner cask wall due to the weight of the internals is represented as a cosine varying pressure around the upper radial portion (105° to 180° range) of the cavity.

In addition, the axial pressure due to the weight of the internals is applied to the cask inner lid surface:

Pia = 22.0 x (66,700) / ((3.14159)*36.02) = 360.4 psi Page 2.10.1-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 C. Pressure Due to Rear Impact Limiter The inertia load of the nonstriking impact limiter is also applied to the cask in two mutually perpendicular directions. The axial component (Pra) is applied as a uniform pressure over the outside surface at the interface with the impact limiter on the bottom end. The pressure applied is calculated as:

Pra = 22.0 x (17,000) / (3.14159)*45.52) = 57.5 psi The radial component (Prr) is assumed to follow a cosine distribution around the lower half of the outside surface (0° to 75° range) of the cask.

D. Reaction Pressures Due to Front Impact Limiter The axial reaction pressure (Pra) was applied over entire interface with the impact limiter on the lid end as shown by the crush footprint of the front impact limiter by Appendix 2.10.8. A total area of 6,503.88 in2 (assuming a 360° arc) was assumed and the pressure applied in two pressure steps as calculated below:

45.5 A2 A1 20 For the axial reaction pressure applied, a total area (6,503.8 in2), and a 360° arc, is calculated as:

A = (() x 45.52) = 6503.8 in2 Axial pressure is applied as two-step (p1 and p2) uniform pressures on total area.

The cask and impact limiter load is applied on area A1 and the internals load is applied on the remaining area, A2.

A1, (as calculated in Load Case IL-16) = 1059.3 in2 A2 = 6503.8 -1059.3 = 5444.5 in2 Page 2.10.1-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Pressure on area A1:

p1 = 22.0 x (169,800+17,000) / 1059.3 = 3,880 psi Pressure on area A2:

p2 = 22.0 x (66,700) / 5444.5 = 270 psi The radial reaction pressures (Prr1 and Prr2) are applied to the crush footprint in the radial direction and assume a cosine varying pressure from 90.5° to 180° of the cask. Since the crush footprint is a circular segment, and the pressures are being applied to two separate surfaces (i.e., lid and flange), the total force, F, applied in the equation is based on the percentage of the total length the specific pressure is applied to.

The calculated reaction axial and radial pressures due to the crushed impact limiter are adjusted and a rotational acceleration 1.113g applied to balance the applied loads. Loading and displacement boundary conditions are shown in Figure 2.10.1-23.

19. 20° Slapdown Impact on Bottom End (IL-19)

For the oblique bottom impact, the cask is inclined at approximately 20º from the horizontal. All the applied loads and reaction forces are transformed into axial and normal components respectively. The axial pressure components due to the internals, front impact limiter and impact reaction are assumed uniformly distributed in two steps. All radial pressure components (i.e. pressure due to internals, front impact limiter and impact reactions) are assumed to have cosine variation over a defined arc length.

The forces acting in this case are:

A. Cask Body Inertia The component accelerations (i.e. 22g axial & 39g Radial - Appendix 2.10.8) are applied as translational inertial loads in the axial and radial directions, respectively. In addition, a rotational acceleration of 1.092g is applied at the vessel C.G. to counteract the out-of-balance and return the model to static equilibrium.

B. Pressure Due to Internals:

Radial pressure (Pir) acting on the lower half of the inner cask wall due to the weight of internals is represented as a cosine varying pressure around the lower radial portion (0° to 75° range) of the cavity.

In addition, an axial pressure due to the weight of the internals is applied to the cask inner lid surface:

Pia = 22.0 x (66,700) / ((3.14159)*36.02) = 360.4 psi Page 2.10.1-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 C. Pressure Due to Front Impact Limiter The inertia load of the nonstriking impact limiter is also applied to the cask in two mutually perpendicular directions. The axial component (Pla) is applied as a uniform pressure over the outer surface of the lid. The pressure applied is calculated as:

Pla = 33.0 x (18,200) / ((3.14159)*45.52) = 61.6 psi The other component (Plr) is assumed to follow a cosine distribution around the upper half of the outside surface (105° to 180° range) of the cask.

D. Reaction Pressures Due to Bottom Impact Limiter The reaction pressure of the striking impact limiter is applied to the cask in two directions (i.e. axial at the cask base and radial at the cask outer wall). The axial component (Pra) is applied as a uniform pressure in two steps over the entire interface with the bottom impact limiter. The crush footprint of the rear impact limiter is shown in Appendix 2.10.8. For the axial reaction pressure applied, a total projected area of 6,503.8 in2 (assuming a 360° arc) is assumed and the pressure is applied at the bottom cask surface along the crush footprint.

For the axial reaction pressure applied, areas A1 and A2 are the same as were calculated the oblique drop on the lid.

A1 = 1,059.3 in2 A2 = 5,444.5 in2 Pressure on area A1:

p1 = 22.0 x (169,800+18,200) / 1059.3 = 3,904.5 psi Pressure on area A2:

p2 = 22.0 x (66,700) / 5444.5 = 270 psi The radial component pressure (Prr) is assumed to follow a cosine distribution around the radial crush footprint from 0° to 89.5° of the cask.

The calculated reaction axial and radial pressures due to the crushed impact limiter are adjusted and a rotational acceleration 1.092g applied to balance the applied loads. Loading and displacement boundary conditions are shown in Figure 2.10.1-24.

Page 2.10.1-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.1.2.1 Pressure Distribution over Contact Area of Cask for Impact Load in Transverse Direction The impact load acting in the transverse direction is applied as a load over the contact area between the impact limiter and the outer surface of the cask. The pressure distribution is assumed to be in the longitudinal direction over the 12 inch long impact limiter contact length and vary with a cosine distribution around the circumference of the cask. For the impact conditions, the angle of contact is dependent upon the amount of crush occurring in the impact limiter. The most severe loads result from impacts on the side of the impact limiter. For these conditions, the contact angle between the impact limiter and the cask outer surface will be approximately 180 degrees. For non-crushing surfaces, a contact angle of 150 degrees (75 degree half angle of contact) is used for the cask impact analysis. The circumferential cosine pressure distribution over a half angle, , is calculated as follows:

Pi = Pmax cos(i / 2) where:

Pi = Pressure load at angle i.

Pmax = Peak pressure load, at point of impact, and i = Angle corresponding to point of interest.

The circumferential pressure distribution is illustrated in the following figure.

R i W

Pi Pma Circumferential Pressure Load Distribution The peak pressure load, Pmax, is determined by setting the integral of the vertical pressure components, Qi, equal to the total transverse impact load, Ft, as follows:

i Ft = Qi LRd i =

Pi cos( i )LRd i = P max cos 2

cos( i )LRd i Page 2.10.1-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 P LR i

= max 2

cos 2

+ i + cos i i d i 2

sin 2 + sin 2

= Pmax LR +

2 + 1 1 2

Rearranging terms gives the peak pressure, Pmax:

1 sin + sin F 2 2 Pmax = t +

LR 2 + 1 1 2

Therefore, the pressure at any circumferential location is given by:

1 sin + sin F

Pi = t 2 + 2 cos i LR 2 2 + 1 1 2

where:

Ft = g W W = Weight of internals or impact limiter g = Acceleration in the transverse direction Therefore, 1

sin + sin Pi = g W 2 + 2 cos i LR 2 2 + 1 1 2

2.10.1.3 ANSYS Analysis Results and Report Methodology ANSYS linear elastic analyses are performed for the above individual load cases.

These individual loads are to be combined and evaluated for normal operating and accident conditions as described in Section 2.10.1.1. A summary of maximum nodal stress intensities in each major cask component under each individual load is presented in Table 2.10.1-2.

Page 2.10.1-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.1.4 Trunnion Local Stress Analysis Due To Lifting Load Method of Analysis 10CFR 71.45(a) [1] requires that any lifting attachment (trunnions) which is a structural part of package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner. The TN-40 trunnion design meets the 10CFR71.45(a) [1] requirements. ASME code allowable is used to evaluate the stress at the trunnion/gamma shell interface. The maximum local membrane (Pl) and local membrane plus secondary (Pl + Q) stress intensities are limited to 1.5 Sm amd 3.0 Sm [4] respectively.

The local stress induced in the Gamma Shield Shell by the trunnions are calculated using Bijlaards method conservatively using 6g. The neutron shield and thin outer shell are not considered to strengthen either the trunnions or the gamma shield shell.

The Trunnion is approximated by an equivalent attachment so that the curves of the Reference WRC-107 [7] can be used to obtain the necessary coefficients. These resulting coefficients are inserted into Table 5 of [7]. The stresses are calculated by performing the indicated multiplication in the column entitled Compute Absolute Values of Stress and Enter Result. The resulting stress is inserted into the stress table at the eight stress locations, i.e., AU, AL, BU, BL, etc. The stresses are calculated by completing Table 5 of Reference [7]. Table 2.10.1-3 contains the computation results for the 6g lifting loads.

Results/Conclusions The maximum stress intensity on the outside and inside of the gamma shield shell for the lifting loads are calculated in Table 2.10.1-3. The maximum stress intensity calculated in Table 2.10.1-3 are combined with the pressure stress and thermal stress calculated in Table 2.10.1-2. The maximum Pl and Pl + Q stress intensities are then listed in the following table and compared with the allowables.

Stress Stress Allowable Allowable Intensity Intensity 1.5 Sm (ksi) 3.0 Sm (ksi)

Pl (ksi) Pl + Q Outside Edge of 12.4 33.75 20.58 67.5 Gamma Shield Shell Inside Edge of 12.4 33.75 19.97 67.5 Gamma Shield Shell Page 2.10.1-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.1.5 References

1. 10CFR Part 71, Packaging and Transportation of Radioactive Material.

2. Roark, Formulas For Stress and Strain, Fourth Edition.

3. Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping casks for Radioactive Material Rev. 1, March 1989.

4. ASME Code Section III, Subsection NB and Appendices, 1989.

5. Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels Rev. 1, March 1978.

6. ANSYS Engineering Analysis System, Users Manual for ANSYS Release 8.0.

7. WRC Bulletin 107, March 1979, Local Stresses in Spherical and Cylindrical Shells Due to External Loadings.

8. Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Rev. 10, October 11, 2005.

9. NUREG 766510, Shock and Vibration Environments for Large Shipping Containers on Rail Cars and Trucks, June 1977.

10. John Harvey, Theory and Design of Modern Pressure Vessels, Second Edition.

Page 2.10.1-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.1-1 Element types, Material Numbers and Real Constants of ANSYS Model Cask Components Component Element Type No. Material No. Real Constant No.

Containment Inner 3 1 -

Shell Containment Bottom 3 1 -

Plate Containment Flange 3 2 -

Lid Outer Plate 6 4 -

Lid Shielding Plate 7 3 -

Bolt Shank 2 5 2 Bolt Head 1 5 3 Bolt Thread 2 5 4 Gamma Shield Shell 4 6 -

Gamma Bottom 5 3 -

Shield Page 2.10.1-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.1-2 Summary Maximum Nodal Stress Intensities in Cask Components for Individual Load Runs (See Figure 2.10.1-3 to Figure 2.10.1-7 for component definition)

Maximum Nodal Stress Intensity (ksi)

Load Case Number Inner Gamma Gamma Lid Flange Shell Shield Shell Bottom Plate IL - 1 8.06 4.90 0.17 0.31 0.02 Bolt Pre-load IL - 2 1.08 13.01 14.45 5.93 3.60 Fabrication IL - 3 2.07 2.21 1.54 1.52 4.05 Int. Press.

IL - 4 0.52 0.55 0.38 0.38 1.02 Ext. Press.

IL - 5 4.02 5.20 4.87 3.21 8.65 Therm. 100F IL - 6 1.31 3.48 3.43 2.29 5.02 Therm. -20F IL - 7 1.26 3.13 3.15 2.29 4.27 Therm.-40F IL - 8 0.26 0.21 0.19 0.32 0.19 Gravity 1g IL - 9 1.42 1.35 1.13 1.98 1.29 Tie-down IL - 10 0.11 0.09 0.07 0.13 0.08 Vibration IL - 11 1.75 1.42 1.18 2.11 1.39 Rail Shock IL - 12 0.12 0.11 0.09 0.08 0.08 End Drop, Bottom 1g IL - 13 0.08 0.14 0.17 0.09 0.18 End Drop, Top, 1g IL - 14 2.81 2.25 1.84 9.58 4.88 Lifting -6g IL - 15 0.53 0.71 0.24 0.83 0.38 Side Drop-1g IL - 16 38.83 31.31 12.68 36.16 12.35 Corner Drop, Lid IL - 17 3.97 3.89 15.22 15.93 15.58 Corner Drop, Bottom IL - 18 49.83 45.24 14.55 60.92 7.49 Oblique Drop, Lid IL - 19 5.05 3.87 16.68 26.04 24.70 Oblique Drop, Bottom Page 2.10.1-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.1-3 6g Lifting Load Trunnion Loading Geometry Cask Loading Cask Weight lb 250000 Gamma Shield Thickness (in) 7.295 Longitudinal 6 Mean Radius (in) 41.150 Vertical g 0 Moment Arm (in) 9.22 Trunnion Outer Radius (in) 6.000 Lateral g 0 Circumferential Trunnion Moment Mc (in lb) 0 Geometry Factor Gamma 5.641 Longitudinal Trunnion Moment ML (in lb) -6915000 Geometry Factor Beta 0.1280 Torsional Trunnion Moment Mt (in lb) 0 P (lb) 0 Circumferential Loading Vc (lb) 0 Longitudinal Loading VL (lb) -750000 Reference Figure Reference Curve from Fig Multiplier Absolute Stress (psi) Au Al Bu Bl Cu Cl Du DI 3c or 4c 0.97 0.000 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1c or 2c-1 0.2 0.000 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 3a 0.07 0.000 0.0 0.0 0.0 0.0 0.0 1a 0.11 0.000 0.0 0.0 0.0 0.0 0.0 3b 0.26 -4373.380 -1137.1 1137.1 1137.1 -1137.1 -1137.1 1b or 1b-1 0.068 -148017.489 -10065.2 10065.2 -10065.2 -10065.2 10065.2 Summation of Phi Stress 11202.3 -8928.1 -11202.3 8928.1 0.0 0.0 0.0 0.0 3c or 4c 0.97 0.000 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1c-1 or 2c 0.2 0.000 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 4a 0.09 0.000 0.0 0.0 0.0 0.0 0.0 2a 0068 0.000 0.0 0.0 0.0 0.0 0.0 4b 0.07 -4373.380 -306.1 306.1 306.1 -306.1 -306.1 2b or 2b-1 0.105 -148017.489 -15541.8 15541.8 -15541.8 -15541.8 15541.8 Summation of Chi Stress 15848.0 -15235.7 -15848.0 15235.7 0.0 0.0 0.0 0.0 Torsional Shear Stress 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Circumferential Shear Stress 0.0 0.0 0.0 0.0 0.0 Longitudinal Shear Stress -5454.3 5454.3 5454.3 -5454.3 -5454.3 Summation of Tau Stress 0.0 0.0 0.0 0.0 5454.3 5454.3 -5454.3 -5454.3 Stress Intensity Root 1 15848.0 8928.1 11202.3 15235.7 5454.3 5454.3 5454.3 5454.3 Stress Intensity Root 2 11202.3 15235.7 15848.0 8928.1 5454.3 5454.3 5454.3 5454.3 Stress Intensity Root 3 4645.7 6307.6 4645.7 6307.6 10908.5 10908.5 10908.5 10908.5 Max Stress Intensity 15848.0 Membrane Stress Intensity Root 1 1137.1 1137.1 306.1 306.1 5454.3 5454.3 5454.3 5454.3 Membrane Stress Intensity Root 2 306.1 306.1 1137.1 1137.1 5454.3 5454.3 5454.3 5454.3 Membrane Stress Intensity Root 3 830.9 830.9 830.9 830.9 10908.5 10908.5 10908.5 10908.5 Max Membrane Stress Intensity 10908.5 Page 2.10.1-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-1 TN-40 Containment Vessel Key Dimensions Page 2.10.1-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-2 TN-40 Cask/Lid/Bolt FEM Representation Page 2.10.1-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-3 Finite Element ModelLid Page 2.10.1-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-4 Finite Element Model - Containment Flange Page 2.10.1-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-5 Finite Element Model - Inner Shell and Bottom Plate Page 2.10.1-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-6 Finite Element Model - Gamma Shield Shell Page 2.10.1-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-7 Finite Element Model - Gamma Bottom Shield Page 2.10.1-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-8 Coupling and Boundary Conditions Page 2.10.1-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-9 Bolt Preload - Loading and Displacement Boundary Conditions Page 2.10.1-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-10 Fabrication Load - Loading and Displacement Boundary Conditions Page 2.10.1-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-11 Internal Pressure - Loading and Displacement Boundary Conditions Page 2.10.1-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-12 Thermal Stress 100 °F Environment - Loading and Displacement Boundary Conditions Page 2.10.1-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-13 Thermal Stress -20 °F Environment - Loading and Displacement Boundary Conditions Page 2.10.1-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-14 Thermal Stress -40 °F Environment - Loading and Displacement Boundary Conditions Page 2.10.1-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-15 1g Horizontal - Loading and Displacement Boundary Conditions Page 2.10.1-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-16 Transport Tie-Down - Loading and Displacement Boundary Conditions Page 2.10.1-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-17 End Drop on Bottom - Loading and Displacement Boundary Conditions Page 2.10.1-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-18 End Drop on Lid - Loading and Displacement Boundary Conditions Page 2.10.1-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-19 6g Vertical Lifting - Loading and Displacement Boundary Conditions Page 2.10.1-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-20 Side Drop, 1g - Loading and Displacement Boundary Conditions Page 2.10.1-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-21 Corner Drop on Lid - Loading and Displacement Boundary Conditions Page 2.10.1-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-22 Corner Drop on Bottom - Loading and Displacement Boundary Conditions Page 2.10.1-49

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-23 20° Slap Down Impact On Lid End -

Loading and Displacement Boundary Conditions Page 2.10.1-50

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.1-24 20° Slap Down Impact on Bottom End-Loading and Displacement Boundary Conditions Page 2.10.1-51

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.2 LID BOLT ANALYSIS TABLE OF CONTENTS 2.10.2.1 Introduction........................................................................ 2.10.2-1 2.10.2.2 Lid Bolt Load Calculations ................................................. 2.10.2-2 2.10.2.3 Summary of Lid Bolt Loads ............................................... 2.10.2-8 2.10.2.4 Lid Bolt Load Combinations............................................... 2.10.2-9 2.10.2.5 Lid Bolt Stress Calculations ............................................. 2.10.2-11 2.10.2.6 Analysis Results .............................................................. 2.10.2-14 2.10.2.7 Lid Bolt Fatigue Analysis ................................................. 2.10.2-14 2.10.2.8 Lid Seal Contact Evaluation ............................................ 2.10.2-18 2.10.2.9 Minimum Engagement Length for Bolt and Flange ......... 2.10.2-19 2.10.2.10 Conclusions ..................................................................... 2.10.2-21 2.10.2.11 References ...................................................................... 2.10.2-22 LIST OF TABLES Table 2.10.2-1 Design Parameters For Lid Bolt Analysis.............................. 2.10.2-23 Table 2.10.2-2 Bolt Data ............................................................................... 2.10.2-24 Table 2.10.2-3 Allowable Stresses In Closure Bolts For Normal Conditions. 2.10.2-24 Table 2.10.2-4 Allowable Stresses In Closure Bolts For Accident Conditions ............................................................................. 2.10.2-25 LIST OF FIGURES Figure 2.10.2-1 Lid/Cask Axial Interface ........................................................ 2.10.2-26 Figure 2.10.2-2 TN-40 Transport Cask (CG Over Corner Lid Drop - Hot)

Seal Decompression as a Function of Circumferential Location ................................................................................ 2.10.2-27 Page 2.10.2-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LID BOLT ANALYSIS 2.10.2.1 Introduction This Appendix evaluates the ability of the cask closure bolt to maintain a leak tight seal under events defined by Normal Conditions Transport (NCT) and the Hypothetical Accident Conditions (HAC). Also evaluated in this section are the bolt thread and internal thread stresses, and lid bolt fatigue. The stress analysis is performed in accordance with NUREG/CR-6007 [1].

The TN-40 cask lid closure arrangement is shown in Figure 2.10.2-1. The 4.5 in. thick lid with a 6.0 in. radiation shield is bolted directly to the shell flange by 48 high strength alloy steel 1.375 in. diameter bolts (with 1 1/2 -8UN threaded portion). Close fitting alignment pins ensure that the lid is centered in the vessel. The bolt material is SA-540 Gr. B23/B24 CL1.

The lid bolt analysis presented in this appendix is done in accordance with NUREG/CR-6007 and conservatively uses lid bolt material of SA-320 Grade L43 with yield strength of 105 ksi and tensile strength of 125 ksi at 70 °F. One exception to this is the minimum engagement length determination in Section 2.10.2.9 where the higher strength bolt material is used. The actual lid bolt material used is SA-540 Grade B23/B24 CL1 with yield strength of 150 ksi and tensile strength of 165 ksi at 70 °F. The lid bolt evaluation due to delayed impact is presented in Appendix 2.10.11 and is based on the lid bolt material of SA-540 Grade B23/B24 CL1.

The following ways to minimize bolt forces and bolt failures for shipping casks are taken directly from Reference [1], page xiii. All of the following design methods are employed in the TN-40 closure system.

Protect closure lid from direct impact to minimize bolt forces generated by free drops (use impact limiters).

Use materials with similar thermal properties for the closure bolts, the lid, and the cask wall to minimize the bolt forces generated by fire accident.

Apply sufficiently large bolt preload to minimize fatigue and loosening of the bolts by vibration.

Lubricate bolt threads to reduce required preload torque and to increase the predictability of the achieved preload.

Use closure lid design which minimizes the prying actions of applied loads.

When choosing a bolt preload, pay special attention to the interactions between the preload and thermal load and between the preload and the prying action.

The following lid bolt evaluations are presented in this section:

Lid bolt torque

Bolt preload Page 2.10.2-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

Gasket seating load

Pressure load

Temperature load

Impact load

Puncture load

Bearing stress

Load combinations for NCT and HAC

Bolt stresses and allowable stresses

Lid bolt fatigue

Lid/Cask seal evaluation

Thread engagement length evaluation The design parameters of the lid closure, taken from Reference [1] are summarized in Table 2.10.2-1. The lid bolt data and material allowables are presented in Table 2.10.2-2 through Table 2.10.2-4. A maximum temperature of 300oF is used in the lid bolt region during NCT and HAC based on results of thermal analyses documented in Chapter 3. The following load cases are considered in the analysis.

1) Preload + Temperature Load (NCT)

2) Pressure Load + 1 Foot Drop (NCT)

3) Pressure + 30 Foot Corner Drop (HAC)

4) Pressure + Puncture Load (HAC) 2.10.2.2 Lid Bolt Load Calculations 2.10.2.2.1 Lid Bolt Torque The desired maximum preload stress in the lid bolts is 50,000 psi.

For a 1.375" bolt shank, the Tensile Stress Area is 1.485 in2 (see Table 2.10.2-2.

Therefore, Fa = 50,000 Stress Area = 50,000 1.485 = 74,250 lb.

The torque required to achieve this preload is (Reference [1], Section 4.0):

Q = K Db Fa = 0.135 (1.375) (74,250) = 13,783 in. lb. = 1,149 ft. lb.

A bolt torque range of 1,100 to 1,150 ft. lb. has been selected. For the minimum torque, Fa = Q/KDb = 1,100 12/(0.135 1.375) = 71,111 lbs Page 2.10.2-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.2.2 Bolt Preload The method used for the following calculation is taken from Reference [1], Table 4.1.

Fa = Q/KDb = 1,150 x 12/0.135(1.375) = 74,343 lb.

Residual torsional moment is:

Mtr = 0.5Q =.5(1,150 x 12) = 6,900 in. lb.

Residual tensile bolt force, Far = Fa = 74,343 lbs 2.10.2.2.3 Gasket Seating Load Gasket characteristics for the Helicoflex HND 229 seals with an aluminum jacket and a 0.236 seal cross section are taken from Reference [2]. The diameter of the inner seal, Dls, is 74.3 in., and the diameter of the outer seal, Dos, is 75.9 in. The force to seat the seals is approximately 1399 lbs./in for aluminum jacket [2] and approximately 2198 lbs/in. for a silver jacket [6]. Therefore the total force required to seat the seals is:

Inner: (74.3) (1399) = 326,555 lbs; (74.3)(2198) = 513,058 lbs Outer: (75.9) (1399) = 333,587 lbs; (75.9)(2198) = 524,106 lbs Total, Fa = 660,142 lbs; Fa = 1,037,164 lbs Therefore, the seal seating load is:

Fa/48 = 660,142/48 = 13,753 lb/bolt; Fa/48 = 1,037,164/48 = 21,608 lb/bolt The specified preload has the required force to seat the seals.

2.10.2.2.4 Pressure Loads The method used for the following calculation is taken from Reference [1], Table 4.3.

Axial force per bolt due to internal pressure is:

Dlg2 (Pli Plo )

Fa =

4Nb Dlg for outer seal (conservative) = 75.9 in. Then, (75.9 2 )(100 0)

Fa = = 9,426 lb./bolt 4( 48 )

Page 2.10.2-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The fixed edge closure lid force is:

Dlb (Pli Plo ) 79 .31(100 )

Ff = = = 1,983 lb. in.-1 4 4 The fixed edge closure lid moment is:

(Pli Plo )Dlb2 100 (79.312 )

Mf = = = 19,656 in. lb. in.-1 32 32 The shear bolt force per bolt is:

El t l (Pli Plo )Dlb (27.8 10 6 )(5.0)(100 )(79.31) 2 2 Fs = = = 32,501 lb./bolt 2Nb E c t c (1 Nul ) ( )

2(48 ) 27.8 10 6 (9.5)(0.7)

The lid shoulder takes this shear force, so that Fs = 0.

2.10.2.2.5 Temperature Loads The lid bolt material is SA-320 Grade L43, 1 3/4 Ni 3/4 Cr 1/4 Mo. This is Group E in the thermal coefficients of expansion tables in Reference [5]. Both the lid and flange are made of SA-350 Gr. LF3, 3 1/2 Ni, which is also Group E. Consequently, the bolts, lid and flange have the same coefficient of thermal expansion (6.9 x 10-6 in/in-oF at 3000F).

Therefore, heating to the maximum isothermal temperature will not generate bolt stress.

2.10.2.2.6 Impact Loads The method used for the following calculation is taken from Reference [1], Table 4.5.

The non-prying tensile bolt force per bolt, Fa, is:

1.34 sin( xi )(LF )(ai )(W l + W c ) 1.34 sin( xi )(1.1)(ai )(82,000 )

Fa = = = 2,518 (ai ) sin( xi ) lb/bolt Nb 48 Note: Wl + Wc is conservatively assumed to be 82,000 lbs. [Actual weights from Table 2-6 are 13,910 lbs for lid and lid bolts, 14,690 lbs for basket, rails, and shims, and 52,000 lbs for fuel assemblies resulting in a total weight of 80,600 lbs.]

The shear bolt force is:

cos( xi )(ai )(Wl ) 82,000(ai ) cos( xi )

Fs = = = 1,708(ai ) cos( xi ) lb/bolt Nb 48 The lid shoulder during normal and accident condition drops takes shear force.

Therefore, Fs = 0 Page 2.10.2-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The fixed-edge closure lid force, Ff, is:

1.34 sin( xi )(LF )(ai )(W l + W c ) 1.34 sin( xi )(1.1)(ai )(82,000 )

Ff = = = 485 .1sin( xi )(ai ) lb/bolt Dlb (79 .31)

The fixed-edge closure lid moment, Mf, is, 1.34 sin( xi )(LF )(ai )( Wl + Wc ) 1.34 sin( xi )(1.1)(ai )(82,000 )

Mf = = = 4,809 sin( xi )(ai ) lb/bolt 8 8 NCT Impact Loads Since the bolts are protected by the impact limiter during an end drop, the worst case scenario is taken to be roughly a 63.8 C.G. over corner drop. From the impact limiter 1 foot normal condition analysis (Appendix 2.10.8), the maximum g load for a 1 foot 63.8 C.G. over corner drop is 5g vertical and 3g horizontal.

However, for the lid bolt analysis the following normal condition g-loading is conservatively used:

ai = 10 gs, and xi = 63.8 Therefore, Fa = 2,518 x 10 x sin(63.8) = 22,594 lb./bolt Fs = 0 lb./bolt Ff = 485.1 x 10 x sin(63.8) = 4,353 lb./bolt Mf = 4,809 x 10 x sin(63.8) = 43,151 lb./bolt HAC Impact Loads The loads resulting from a 30 foot, 63.8 corner drop are conservatively taken to be the following (actual loads are 32 g vertical, and 14 g horizontal, given in Appendix 2.10.8).

ai = 38 g, and xi = 63.8 Therefore, Fa = 2,518 x 38 x sin(63.8) = 85,856 lb./bolt Fs = 0 lb./bolt Ff = 485.1 x 38 x sin(63.8) = 16,540 lb./bolt Mf = 4,809 x 38 x sin(63.8) = 163,973 lb./bolt Page 2.10.2-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.2.7 Puncture Loads The method used for the following calculation is taken from Reference [1], Table 4.7.

The non-prying tensile bolt force per bolt, Fa, is:

sin( xi )Pun Fa =

Nb where:

0.75D pb2 S yl Pun = The smaller of 0.6D pb t l Sul 0.75 (6 2 )(33,200) = 2.816 10 6

= The smaller of 6 0.6 (6)(10.5)(70,000) = 8.31 10 pun = 2.816x106 lb The puncture force is greatest when xi = 90. Conservatively neglect the protection provided by the impact limiter. Then, sin( xi )2.816 10 6 Fa = = 58,669 lb 48 Since this force is negative (inward acting), the actual resulting bolt force, Fa = 0, because the applied load is supported by the cask wall and not the lid bolts. The shear bolt force is:

cos(90 )Pun Fs = lb/bolt Nb The lid shoulder during puncture takes shear force. Therefore, Fs = 0 The fixed-edge closure lid force, Ff, is:

sin( xi )Pun sin(90 )2.816 10 6 Ff = = = 11,302 lb/bolt Dlb (79.31)

The fixed-edge closure lid moment, Mf, is, sin( xi )Pun sin(90 )2.816 10 6 Mf = = = 224,100 lb/bolt 4 4 Page 2.10.2-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.2.8 External Pressure Load of 290 psig An external pressure load of 290 psig is evaluated as shown in Reference [1], Table 4.3.

The axial force per bolt due to internal pressure is, Dlg2 (Pli Plo )

Fa =

4N b Where the outer seal diameter, Dlg, is Dlg = 75.9 in.

Then, (75.9 2 )(0 290)

Fa = = 27,336 lb./bolt 4( 48)

Since this force is negative (inward acting), the actual resulting bolt force, Fa = 0, because the applied load is supported by the cask wall and not the lid bolts.

The fixed edge closure lid force Ff is, Dlb (Pli Plo ) 79.31( 290)

Ff = = = 5,750 lb in.-1 4 4 The fixed edge closure lid moment, Mf, is, (Pli Plo )Dlb2 290(79.312 )

Mf = = = 57,004 in. lb in.-1 32 32 The lid shoulder during external pressure takes shear force. Therefore, Fs = 0 Page 2.10.2-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.3 Summary of Lid Bolt Loads The loads calculated in the previous sections are summarized in the following table.

LID BOLT INDIVIDUAL LOAD

SUMMARY

Non-Prying Torsional Prying Moment, Load Applied Prying Force, Tensile Force, Fa Moment, Mf Case Load Ff (lb.in.-1)

(lb.) Mt (in. lb.) (in. lb. in.-1)

Maximum 74,343 6,900 0 0 Torque Preload Residual Minimum 71,111 6,600 0 0 Torque Gasket Seating Load 21,608 0 0 0 Pressure 100 psig Internal 9,426 0 1,983 19,656 Thermal 300F 0 0 0 0 1 Foot Normal Condition Drop 22,594 0 4,353 43,151 Impact 30 foot Accident Condition 85,856 0 16,540 163,973 Drop Puncture Drop on six inch diameter rod 0 0 -11,302 -224,100 External 290 psig 0 0 -5,750 -57,004 Pressure External Page 2.10.2-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.4 Lid Bolt Load Combinations A summary of normal and accident condition load combinations is presented in the following table. The method used for the following combination is taken from Reference

[1], Table 4.9.

LID BOLT NORMAL AND ACCIDENT LOAD COMBINATIONS Non-Prying Torsional Prying Load Prying Force, Combination Description Tensile Force, Fa Moment, Moment, Mf Case Ff (lb.in.-1)

(lb.) Mt (in. lb.) (in. lb. in.-1)

A.

Maximum 74,343 6,900 0 0 Preload + Torque Temperature 1

(Normal Condition) B. Minimum 71,111 6,600 0 0 Torque Pressure + Normal Impact 2 32,020 0 6,336 62,807 (Normal Condition)

Pressure + Accident Impact 3 95,282 0 18,523 183,629 (Accident Condition)

Pressure + Puncture 4 9,426 0 -9,319 -204,444 (Accident Condition) 5 Internal & External Pressure 9,426 0 -3,767 37,348 Additional Prying Bolt Force Since the prying forces applied in load cases 4 and 5 acts inward, normal to the cask lid, an additional prying bolt force, Fap, is generated (Reference [1], Table 2.1). No additional force is generated for the outward loadings however (load cases 1, 2, and 3),

because of the gap between the lid and flange at the outer edge. Only load case 4 is considered because it bounds load case 5, above. Fap is calculated in the following way.

2M f C1(B Ff ) C2 (B P )

D (D Dlb )

Fap = lb li ,

Nb C1 + C2 where, C1 = 1, C2 =

8 E t 3 l l

+

(Dlo Dli )Elf t lf3 Lb 2

( )

3 Dli Dlb Nul 1 Dlb 2 N b Db E b 8 26.7 10 6 (10.5 3 ) (82.75 72.87)(26.7 10 6 )( 4.5)3 4.5

= +

3(72.87 79.31)2 1 0.3 79.31 2 6 (48)(1.375 )(26.7 10 )

= 5.309, Page 2.10.2-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 B is the non-prying tensile bolt force, and P is the bolt preload. Since Fs = 0, Fs < P, and therefore B = P. Parameters B, P, Ff, and Mf are quantities per unit length of bolt circle.

Also, the Pressure load is not included because it decreases the magnitude of the applied prying moment, which is less conservative. For the applied inward force, Fa N b (74,343 )( 48 )

P =B= = = 14,322 lb. in.-1 Dlb (79 .31)

( 224,100)( 48)

Mf = = 224,100 lb. in.-1, and Ff = 0 lb. in.-1 (79.31)

Therefore, 2( 224,100) 1(14,322 0) 5.309(14,322 14,322)

(79.31) (72.81 79.31)

Fap =

48 1 + 5.309

= 45,478 lb/bolt It is observed that the additional tensile bolt force due to prying for the puncture is less than the accident impact force. The puncture is therefore not critical for bolt stress evaluation.

Bolt Bending Moment The method used for the following calculation is taken from Reference [1], Table 2.2.

The maximum bolt bending moment, Mbb, generated by the applied load is evaluated as follows:

D K b Mbb = lb Mf Nb K b + K l The terms Kb and Kl are based on geometry and material properties and are defined in Reference [1], Table 2.2. By substituting the values given above, N E D 4 48 26.7 106 1.3754 64 = 2.006 x 10 5

K b = b b b =

L D b lb 64 4.5 79 .31 E l t l3 26.7 10 6 (10.5 3 )

Kl =

2 2

( )

3 1 Nul + (1 Nul )

2 2 Dlb Dlb ( )

3 1 0.3 2 + (1 0.3) 2 79.31 79.31 Dlo 82.75

= 9.551 x 107 Page 2.10.2-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Therefore, 79.31 2.006 105 M bb = 5 M = 0.0109 Mf 7 f 48 2.006 10 + 9.551 10 For load case 2, Mf = 62,807 in. lb. Substituting this value into the equation above gives, Mbb = 684.6 in. lb/ bolt 2.10.2.5 Lid Bolt Stress Calculations The method used for the following calculation is taken from Reference [1], Table 5.1.

2.10.2.5.1 Average Tensile Stress The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive) force on the closure joint, under both normal and accident conditions. Based upon the load combination it is shown that a positive (compressive) load is maintained on the clamped joint for all load combinations except for the accident condition impact plus pressure load. A more detailed finite element analysis is performed in Section 2.10.2.8 of this Appendix to evaluate closure of the lid during this event. The maximum non-prying tensile force for normal conditions is 74,343 lb, from load case 1.A. (maximum torque preload + temperature load), and the maximum non-prying tensile force for accident conditions is 95,282 lb from load case 3 (accident impact + pressure load). These loads are used to compute bolt stresses below.

NCT:

Fa 74,343 Sba = 1.2732 2

= 1.2732 = 50,065 psi = 50.1 ksi Dba 1.3752 HAC:

95,282 Sba = 1.2732 = 64,166 psi = 64.2 ksi 1.3752 2.10.2.5.2 Bending Stress Normal Condition:

M bb 684.6 Sbb = 10.186 3

= 10.186 = 2,682 psi = 2.7 ksi Dba 1.3753 2.10.2.5.3 Shear Stress For both normal and accident conditions, the average shear stress caused by shear bolt force Fs is, Page 2.10.2-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Sbs = 0 For normal and accident conditions the maximum shear stress caused by the torsional moment Mt is, Mt 6,900 Sbt = 5.093 3

= 5.093 = 13,518 psi = 13.5 ksi Dba 1.3753 2.10.2.5.4 Maximum Combined Stress Intensity The maximum combined stress intensity is calculated in the following way (Reference

[1], Table 5.1).

Sbi = [(Sba + Sbb)2 + 4(Sbs + Sbt)2]0.5 For normal conditions, the combined tension, shear, bending, and residual torsion results in a maximum stress intensity of:

Sbi = [(50,065 + 2,682)2 + 4 (0 + 13,518)2]0.5 = 59,272 psi = 59.3 ksi 2.10.2.5.5 Stress Ratios In order to meet the stress ratio requirement, the following relationship must hold for both normal and accident conditions.

Rt2 + Rs2 < 1 Where Rt is the ratio of average tensile stress to allowable average tensile stress, and Rs is the ratio of average shear stress to allowable average shear stress.

For NCT:

Rt = 50,065/63,800 = 0.785 Rs = 13,518/38,280 = 0.353 Rt2 + Rs2 = (0.785)2 + (0.353)2 = 0.740 < 1 For HAC:

Rt = 64,166/87,500 = 0.733 Rs = 13,518/52,500 = 0.257 Rt2 + Rs2 = (0.733)2 + (0.257)2 = 0.603 < 1 2.10.2.5.6 Bearing Stress Under Lid Bolt Head The maximum NCT axial force is 74,343 lb. A bolt hole of 1.625 diameter is used for the shank (Chapter 1, Appendix 1.4 drawings).

Page 2.10.2-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 H = 2.25/2 in. Diam = 1.625 in B = 1.125 tan(30) = 0.650 in.

Total Area of one triangle

= (1.125)(.650) = .731 in.2 H

Total area under Bolt head - Bolt Hole area

= 6(.731) - (/4)(1.6252) = 2.31 in.2 B The total bearing area is 2.31 in.2 The bearing stress for normal conditions is, Bearing Stress = 74,343/2.31 = 32,179 psi = 32.2 ksi The allowable normal condition bearing stress on the lid is taken to be the yield stress of the lid material at 300°F. The lid is manufactured out of SA-350 Grade LF3, which has a yield stress of 33.2 ksi at 300°F.

Page 2.10.2-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.6 Analysis Results A summary of the lid bolt stresses calculated above is presented in the following table:

SUMMARY

OF STRESSES AND ALLOWABLES NCT HAC Stress Type Stress Allowable Stress Allowable Average Tensile (ksi) 50.1 63.8 64.2 87.5 Shear (ksi) 13.5 38.3 13.5 52.5 Not Required Combined (ksi) 59.3 86.1 Reference [1]

Interaction E.Q.

0.740 1 0.603 1 Rt2 + Rs2 < 1 Bearing Allowable (ksi) Not Required 32.2 33.2 (Sy of lid material) Reference [1]

The calculated bolt stresses are all less than the specified allowable stresses.

2.10.2.7 Lid Bolt Fatigue Analysis The purpose of the fatigue analysis is to show quantitatively that the fatigue damage to the bolts during NCT is acceptable. This is done by determining the fatigue damage factor for each NCT event. For this analysis it is assumed that the bolts are replaced after 50 round trip shipments. The total cumulative damage or fatigue usage for all events was conservatively determined by adding the usage factors for the individual events. The sum of the individual usage factors was checked to make certain that for the 50 round trip shipments of the TN-40 cask, the total usage factor was less than one.

The following sequence of events was assumed for the fatigue evaluation.

1) Operating Preload (Bolt Tensile stress, Sba = 50,000 psi, and bolt torsional shear stress, Sbt = 13,518 psi, corresponding to a bolt torque of 1,150 ft. lb.), with 50 round trip shipments considered.

2) Test pressure

3) Rail vibration / shock

4) Pressure and temperature fluctuations

5) 1 foot normal condition drop Since the bolt preload stress applied to the TN-40 cask lid bolts is higher than all of the other NCT condition loads, the stress in the bolt will never exceed the bolt preload stress. Consequently, the application and removal of preload is the only real cyclic loading that occurs in the lid bolts. The following analysis is therefore very conservative since it assumes that the damage factor is the sum of all of the individual event damage factors, and not simply the damage factor for bolt preload.

2.10.2.7.1 Operating Preload Assuming that the bolts are replaced after 50 round trips, the number of preload cycles is two times the number of trips or 100 cycles.

Page 2.10.2-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The maximum normal condition bolt stress intensity is 59.3 ksi (Section 2.10.2.5.4).

2.10.2.7.2 Test Pressure The ASME code-mandated proof test [10] is 1.25 (Design Pressure) = 125 psi, and will only be performed once.

From Section 2.10.2.2, the 100 psi internal pressure load analysis can be used by scaling the results upward by a factor of 1.25.

Fa = 9,4261.25 = 11,783 lb / bolt, Fs = 01.25 = 0 lb / bolt Ff = 1,9831.25 = 2,479 lb / in.

Mf = 19,6561.25 = 24,570 in-lb / in.

Mbb = 0.0109 Mf = 267.8 in-lb / bolt The lid bolt diameter is 1.375 in. Therefore from Reference [1], we get the following:

Fa 11,783 Sba = 1.2732 2

= 1.2732 = 7,935 psi Dba 1.3752 M bb 267 .8 Sbb = 10 .186 3

= 10 .186 = 1,049 psi Dba 1.375 3 Since internal pressure causes no bolt torsion, and all shear loads are taken by the lid

shoulder, Shear stress:

Sbs = 0, and Sbt = 0.

Stress Intensity:

S.I. = Sbi = [(Sba + Sbb)2 + 4(Sbs + Sbt)2]0.5 = [(7,935 + 1,049)2 + 4(0)2]0.5 =

8,984 psi Page 2.10.2-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.7.3 Vibration / Shock Since the TN-40 Package will be shipped by rail car, the shock and vibration loadings for rail configurations only will be considered.

Rail Car Shock:

Again, assume 50 round trip shipments, averaging 2000 miles each way. Reference [4]

reports that there are roughly 9 shock cycles per 100 miles of rail car transport.

Therefore the total number of cycles is 2000 miles 2 round trip 150 shipments 0.09 shocks per mile = 18,000 cycles.

Reference [4] also specifies a peak shock loading of 4.7 g in the longitudinal direction for rail car transport. Consequently, the bolt force due to rail car shock is (82,000 lb)(4.7 g) / (48 bolts)(1.485 in2 per bolt) = 5,407 psi Vibration:

According to Reference [4], the peak vibration load on the deck of a rail car in the longitudinal direction is 0.19 g. This results in a stress of 219 psi, which is negligible for a high strength bolt.

2.10.2.7.4 Pressure and Temperature Fluctuations The lid bolt material is SA320 GR. L43, 13/4Ni 3/4Cr 1/4Mo, which is in group E in the coefficients of thermal expansion tables in Reference [5]. The lid and flange are both made of SA 350 GR LF3, 3 1/2 Ni, which is also group E. Therefore the lid bolts and all of the materials it contacts have the same coefficient of thermal expansion. Consequently thermal load will cause no stress in the lid bolts.

The pressure fluctuation is conservatively assumed to be the maximum design pressure, 100 psi, which is far greater that the actual operating pressure. Since the stress intensity in the lid bolts is linearly proportional to the internal / external pressure difference, the stress intensity due to 100 psi internal load is, 100psi 8,984 psi = 7,187 psi 125psi The pressure fluctuation is assumed to occur once per round trip, since there is no payload, and therefore no pressurization, during the return trip. So the total number of cycles of pressure fluctuation is 50.

Page 2.10.2-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.7.5 1 Foot Normal Condition Drop The normal condition drop consists of a 1 foot drop in an orientation that results in the most damage. For the side drop the resulting shear load is taken entirely by the lid /

flange interface. For the end drop, the load is transferred to the cask body via the impact limiters, protecting the bolts. Therefore the worst case scenario is taken to be roughly a 63.8 C.G. over corner drop.

The lid bolt analysis above conservatively uses a 63.8° corner drop axial acceleration of 10 g to calculate the following bolt loads:

Fa = 22,594 lb/ bolt Fs = 0 lb./bolt Ff = 4,353 lb./bolt and Mf = 43,151 lb./bolt Mbb = 0.0109 Mf = 470 in-lb / bolt The lid bolt diameter is 1.375 in. Therefore, from Reference [1], we get the following Fa 22,594 Sba = 1.2732 2

= 1.2732 = 15,215 psi Dba 1.3752 M bb 470 Sbb = 10.186 3

= 10.186 = 1,842 psi Dba 1.3753 Since internal pressure causes no bolt torsion, and all shear loads are taken by the lid

shoulder, Shear stress Sbs = 0, and Sbt = 0.

Stress intensity S.I. = Sbi = [(Sba + Sbb)2 + 4(Sbs + Sbt)2]0.5

= [(15,215 + 1,842)2 + 4(0)2]0.5 = 17,057 psi Conservatively assume that the cask is dropped once per shipment, resulting in 50 normal condition drops before the lid bolts are changed.

Page 2.10.2-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.7.6 Damage Factor Calculation The following damage factors are computed based on the stresses and cyclic histories described above, a fatigue strength reduction factor, KF, of 4 (Reference [5]), and the fatigue curve shown in Table I-9.4 of Reference [5].

Stress Cycles Damage S.I. KF Event Intensity Sa (psi) Factor (psi) (psi) n N n/N Operating Preload 59,300 237,200 128,088 100 619 0.16 Test Pressure 8,984 35,936 19,405 1 81,054 0.00 Rail Car Shock 5,407 21,628 23,354 18,000 35,314 0.51 Pressure and 7,187 28,748 15,524 50 106 0.00 Temperature 1 Foot Normal 17,057 68,228 36,843 50 7,701 0.0065 Condition Drop 0.68 Here, n is the number of cycles, N is taken from Figure I-9.4 of Reference [5], and Sa is defined in the following way:

If one cycle goes from 0 to +S.I., then Sa = (1/2) S.I. KF KE.

If one cycle goes from -S.I. to + S.I., then Sa = S.I. KF KE.

Where KE is the correction factor for modulus of elasticity, 30106 / 27.8106 = 1.08 (Reference [5]).

Since the total damage factor is less than one in both cases, the TN-40 cask lid bolts will not fail due to fatigue in either case.

2.10.2.8 Lid Seal Contact Evaluation The lid seal design was conducted in order to determine the lid/cask seal status when subject to a CG over lid corner impact 30 foot drop.

Section 2.10.2.3 above shows that during accident condition the preload is not enough to maintain a compressive force on the seal. An elastic finite element analysis is performed to determine the status of the lid/cask seal during a CG over corner 30 foot drop. The finite element model from Appendix 2.10.1 is modified to include contact elements (CONTAC52) at the lid/cask axial interface, internal pressure, bolt preload, seal load and 30 foot drop conditions.

Page 2.10.2-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.8.1 Assumptions

CG over corner lid impact with internal pressure is the worst case condition.

The force to seat the seals is 1399 lbs./in [2] and 2198 lbs/in [6]. The total load to seat the seal is 660,142 lbs or 1,037,164 lbs, but 700,000 lbs and 1,040,000 lbs will be used conservatively.

The maximum allowable decompression of the seal is 0.040 [2].

2.10.2.8.2 Analysis The finite element model from Appendix 2.10.1 is modified to include contact elements (CONTAC52) at the lid/cask axial interface, internal pressure, bolt preload, seal load and 30 foot drop conditions.

Gap elements (CONTAC52) were used to model the lid/cask axial interface. To get an accurate contact representation a 60 mil axial gap was included radially outwards of

Ø77.25 (closest node at Ø78.10) between the lid/cask axial interface. Figure 2.10.2-1 shows the lid/cask axial interface.

A pressure of 100 psi was applied to all internal surfaces. Bolt shank prestrain was calculated based on = /E, where is the bolt prestress (50 ksi per Section 2.10.2.2 above). The seal loads of 700,000 lbs and 1,040,000 lbs were applied via CONTAC52 elements. The stiffness for the gap element was calculated based on F=kx. The accident drop conditions were kept consistent with Appendix 2.10.1.

2.10.2.8.3 Results Figure 2.10.2-2 plots the decompression of the seal as a function of circumferential location. The maximum decompression is 0.003 in. which is less then the allowable seal decompression of 0.040 in.

From the analysis results presented in the Figures and discussion, it can be concluded that during the CG over corner drop lid impact loading with internal pressure, the metal-to-metal contact exists at the Helicoflex seal. Since a seal exists around the circumference of the TN-40 vessel, the internal contents will not leak during a worst case loading condition.

2.10.2.9 Minimum Engagement Length for Bolt and Flange For a 11/2 - 8UN bolt, the material is SA-540 GR B23/B24 CL1, with Su = 165 ksi, and Sy = 150 ksi (at room temperature)

Page 2.10.2-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The flange material is SA - 350 GR LF3, with Su = 70 ksi, and Sy = 37.5 ksi (at room temperature)

The minimum engagement length, Le, for the bolt and flange is (Reference [7], Page 1490),

2At Le =

1 3.146K n max + .57735n(E s min K n max )

2 where, At = tensile stress area = 1.485 in.2 n = number of threads per inch = 8 Kn max = maximum minor diameter of internal threads = 1.390 in.

Es min = minimum pitch diameter of external threads = 1.4166 in.

Ds min = minimum major diameter of external threads = 1.4978 in.

Substituting the values given above, 2(1.485)

Le = = 1.09 in.

1 (3.1416)1.390 + .57735(8)(1.4166 1.390) 2 As Sue J= , (Reference [7])

An Sui Where, Sue is the tensile strength of external thread material, and Sui is the tensile strength of internal thread material.

As = shear area of external threads

= 3.1416 n Le Kn max [1/(2n) + .57735 (Es min - Kn max)]

An = shear area of internal threads

= 3.1416 n Le Ds min [1/(2n) + .57735(Ds min - En max)]

For the bolt / flange insert connection:

En max = maximum pitch diameter of internal threads = 1.4283 in.

Therefore, As = 3.1416 (8) (1.09) (1.390) [1 / (2 x 8) + .57735 (1.4166 - 1.390)]

= 2.96 in.2 Page 2.10.2-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 An = 3.1416 (8) (1.09) (1.4978) [1 / (2 x 8) + .57735 (1.4978 - 1.4283)]

= 4.21 in.2 So, 2.96(165.0)

J= = 1.66 4.21(70.0)

Therefore, the minimum required engagement length, Q = J Le = 1.66 x 1.09 = 1.81 in.

The actual minimum engagement length

= (6.50 bolt length - 4.50 lid thickness) = 2.00 in. > 1.81 in.

The above calculation bounds the minimum required engagement length if inserts are used because Su of inserts is higher than the Su for the lid thus lowering the J value.

2.10.2.10 Conclusions

A lid bolt torque range of 1,100 to 1,150 ft. lb. is recommended to achieve the desired preload stress of 50,000 psi.

Lid bolt stresses meet the acceptance criteria of NUREG/CR-6007 "Stress Analysis of Closure Bolts for Shipping Casks" [1].

For the recommended preload, a positive (compressive) load is maintained during all load combinations, except for the accident condition impact plus pressure load case.

Closure of the TN-40 Cask lid is evaluated in Section 2.10.2.8 above and the seal remains closed during a worst case impact.

The bolt and flange thread engagement length is acceptable.

Page 2.10.2-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.2.11 References

1. Stress Analysis of Closure Bolts for Shipping Cask, NUREG/CR-6007, 1992.

2. High Performance Sealing, Metal Seals Helicoflex Catalog, Helicoflex Co., Boonton, N.J., ET 507 E 5930.

3. Draft American Standard Design Basis for Resistance to Shock and Vibration of Radioactive Material Packages Greater than One Ton in Truck Transport, ANSI N14.23.

4. Shock and Vibration Environments for Large Shipping Containers on Rail Cars and Trucks, NUREG 766510.

5. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III and Appendix, 1989.

6. Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Rev. 10, 10/11/2005.

7. Machinery Handbook, 26st Ed, Industrial Press, 2000.

8. Baumeister, T., Marks, L. S., Standard Handbook for Mechanical Engineers, 7th Edition, McGraw-Hill, 1967.

9. Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, U.

S. Nuclear Regulatory Commission, Regulatory Guide 7.6, Revision 1, March 1978.

10. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB6220, 1989.

Page 2.10.2-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.2-1 Design Parameters For Lid Bolt Analysis

Db Nominal diameter of closure bolt; 1.375 in.

K Nut factor for empirical relation between the applied torque and achieved preload is 0.135 for neolube

Q Applied torque for the preload (in.-lb.)

Dlb Closure lid diameter at bolt circle, 79.31 in.

Dlg Closure lid diameter at the seal (outer), 75.9 in.

Ec Youngs modulus of cask wall material (SA-350, LF3, 300°F), 26.7106 psi

El Youngs modulus of lid material (SA-350, LF3, 300°F), 26.7 x 106 psi

Nb Total number of closure bolts, 48

Nul Poissons ratio of closure lid, 0.3, (Reference [8], p. 5-6 use nominal\ value).

Pei Inside pressure of cask, 100 psig.

Dlo Closure lid diameter at outer edge, 82.75 in.

Pli Pressure inside the closure lid, 100 psig.

tc Thickness of cask wall, 8.0 + 1.5 = 9.5 in.

tl Thickness of lid center, 10.5 in; lid flange, 4.5 in.

lb Thermal coefficient of expansion, bolt (SA-320, L43), 6.4 x 10-6 at R.T.,

6.9 x 10-6in. in.-1 F-1 at 300F

lc Thermal coefficient of expansion, cask (SA-350, LF3) 6.4 x 10-6 at R.T.,

6.9 x 10-6in. in.-1 F-1 at 300F

ll Thermal coefficient of expansion, lid (SA-350, LF3) 6.4 x 10-6 R.T.,

6.9 x 10-6 in. in.-1 F-1 at 300F

Eb Young's modulus of bolt material (SA-320, L43, 300° F), 26.7 x 106 psi

ai Maximum rigid-body impact acceleration (g) of the cask

LF Load Factor to account for any difference between the rigid body acceleration and the acceleration of the contents and closure lid = 1.1

Wc Weight of contents = 52,000 (fuel) + 14,693 (basket)** = 66,693 lbs.

Wl Weight of lid = 13,907 lbs.

Wc+Wl 66,693 + 13,907 = 80,600 lbs., assume 82,000 lbs.

xi Impact angle between the cask axis and target surface

Syl Yield strength of closure lid material (SA-350, LF3, 300° F), 33,200 psi

Sul Ultimate strength of closure lid (SA-350, LF3, 300° F), 70,000 psi

Syb Yield strength of bolt material (see Table 2.10.2-3

Sub Ultimate strength of bolt material (see Table 2.10.2-4

Plo Pressure outside the lid

Lb Bolt length between the top and bottom surfaces of closure, 4.5 in.

Pun Maximum impact force generated by the puncture bar during a normal impact Dpb Puncture bar diameter, 6 inches per 10 CFR 71.73 (c) (3)

- Conservatively using higher basket weight for lid bolt analysis Page 2.10.2-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.2-2 Bolt Data Parameters necessary to use formulas of Reference [1], Table 5.1.

Bolt: 1 1/2"- UN8 - 2A N: no of threads per inch = 8 p: Pitch = 1/8" = .125 in.

Db: Nominal Diameter = 1.5 in.

Dba: Bolt diameter for stress calculations in the threaded area

= Db - .9743p = 1.5 - .9743 (.125) = 1.378 in Bolt Thread Stress Area = /4 (1.378)2 = 1.491in2 Bolt Shank Stress Area = /4 (1.375)2 = 1.484 in2 Table 2.10.2-3 Allowable Stresses In Closure Bolts For Normal Conditions (MATERIAL: SA-320 Gr. L43)

Normal Condition Allowables Temperature Yield Stress 1

(°F) (ksi) Ftb 2,4 Fvb 3.4 S.I. 5 (ksi) (ksi) (ksi) 100 105.0 70.0 42.0 94.5 200 99.0 66.0 39.6 89.1 300 95.7 63.8 38.3 86.1 400 91.8 61.2 36.7 82.6 500 88.5 59.0 35.4 79.7 600 84.3 56.2 33.7 75.9 Notes:

1. Yield stress values are from [5]

2. Allowable Tensile stress, Ftb = 2/3 Sy [1]

3. Allowable shear stress, Fvb = 0.4 Sy [1]

4. Tension and shear stresses must be combined using the following interaction equation:

2 tb2 yb

+ 1.0 [1]

Ftb2 Fyb2

5. Stress intensity from combined tensile, shear and residual torsion loads, S.I. 0.9 Sy [1]

Page 2.10.2-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.2-4 Allowable Stresses In Closure Bolts For Accident Conditions (MATERIAL: SA-320 Gr. L43)

Accident Condition Allowables Temperature Yield Stress 1

(°F) (ksi) 0.6 Sy 3 Ftb 2,4 Fvb 3,4 (ksi) (ksi) (ksi) 100 105.0 63.0 87.5 52.5 200 99.0 59.4 87.5 52.5 300 95.7 57.4 87.5 52.5 400 91.8 55.1 87.5 52.5 500 88.5 53.1 87.5 52.5 600 84.3 50.6 84.3 50.6 Notes:

1. Yield and tensile stress values are from [5], Note that Su is 125 ksi at all temperatures of interest.

2. Allowable Tensile stress, Ftb = MINIMUM(0.7 Su, Sy), where 0.7 Su = 0.7 (125) = 87.5 ksi [1].

3. Allowable shear stress, Fvb = MINIMUM(0.42 Su, 0.6 Sy), where 0.42 Su = 0.42 (125.) = 52.5 ksi [1].

4. Tension and shear stresses must be combined using the following interaction equation:

2 tb2 yb

+ 1.0 [1]

Ftb2 Fyb2 Page 2.10.2-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.2-1 Lid/Cask Axial Interface Page 2.10.2-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.2-2 TN-40 Transport Cask (CG Over Corner Lid Drop - Hot)

Seal Decompression as a Function of Circumferential Location Page 2.10.2-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.3 STRUCTURAL ANALYSIS OF THE OUTER SHELL TABLE OF CONTENTS 2.10.3.1 Introduction........................................................................ 2.10.3-1 2.10.3.2 Description ........................................................................ 2.10.3-1 2.10.3.3 Materials Input Data .......................................................... 2.10.3-1 2.10.3.4 Applied Loads .................................................................... 2.10.3-1 2.10.3.5 Method of Analysis ............................................................ 2.10.3-2 2.10.3.6 Analysis Results ................................................................ 2.10.3-4 2.10.3.7 References ........................................................................ 2.10.3-6 LIST OF FIGURES Figure 2.10.3-1 Cask Outer Shell and Connection to Cask Body .................... 2.10.3-7 Figure 2.10.3-2 Finite Element Model - Boundary Conditions .......................... 2.10.3-8 Page 2.10.3-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 STRUCTURAL ANALYSIS OF THE OUTER SHELL 2.10.3.1 Introduction This section presents the structural analysis of the outer shell of the TN-40 package.

The outer shell consists of a cylindrical shell section and closure plates at each end which connect the shell to the cask body. The shell is evaluated for Normal Condition of Transport (NCT) that includes internal pressure, normal handling/tiedown loads and 1 foot end/side drops. Maximum membrane and membrane plus bending stress intensities due to the pressure difference, handling/tiedown loads and 1 foot end/side drop loads are determined. These stresses are compared to the allowable stress limits in Chapter 2 to assure that the design criteria are met.

2.10.3.2 Description The outer shell is constructed from low-alloy carbon steel and is welded to the outer surface of the gamma shield shell. The cylindrical shell section is 0.5 in. thick and the closure plates are 0.75 inches thick. Pertinent dimensions are shown in Figure 2.10.3-1 and Drawing 10421-71-3 in Appendix 1.4.1.

2.10.3.3 Materials Input Data The outer shell cylindrical section and closure plates are SA-516 Gr 55. The material properties are taken from the ASME Code [1]. The yield strength of the material is also obtained from the ASME Code [1] at a temperature of 250F. The temperature is a conservative value as compared to the calculated maximum temperature of 214°F given in Chapter 3, Table 3-1.

2.10.3.4 Applied Loads It is assumed that a pressure of 25 psig may be applied to the inside of the outer shell during NCT. However, the external pressure load on the shell is reacted by the resin-filled aluminum containers inside the shell and thus does not create significant stresses in the shell.

The handling loads acting on the outer shell are a result of lifting. The lifting load includes a 3g factor as specified in 10CFR71 [2]. The weight or inertia g load includes the weight of the outer shell, neutron resin, and aluminum containers. The 2g vertical, 5g lateral and 10g longitudinal acceleration tiedown loads are applied when the cask is oriented horizontally to ensure it is not damaged during transport. The most severe NCT loads are the 1 foot end drop and side drops. The drop accelerations are 12g and 16g for the NCT end drop and side drop, respectively. These values are taken from Table 2.10.8-7. The load cases considered consist of the following:

Page 2.10.3-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

Cask in the Vertical Orientation Stress due to 25 psig internal pressure Stress due to 25 psig internal pressure and 3g inertia load (lifting)

Stress due to 25 psig internal pressure and 1 foot end drop (12g)

Cask in the Horizontal Orientation Stress due to 25 psig internal pressure Stress due to 25 psig internal pressure, 2g vertical, 5g lateral and 10g longitudinal forward acceleration Stress due to 25 psig internal pressure, 2g vertical, 5g lateral and 10g longitudinal backward acceleration Stress due to 25 psig internal pressure and 1 foot side drop (16g) 2.10.3.5 Method of Analysis ANSYS Model A finite element model is built for the structural analysis of the outer shell and closure plates. The outer shell and closure plates are modeled with ANSYS Solid 45 elements

[3]. The basic geometry of the outer shell and weld sizes used for analysis are shown in Figure 2.10.3-1. The finite element model is shown in Figure 2.10.3-2.

Cask in the Vertical Orientation

Stress due to 25 psig internal pressure An internal pressure of 25 psig is used as the maximum pressure acting on the inner surface of the outer shell. The maximum shell stress intensity for this load case is 5.02 ksi.

Stress due to 25 psig internal pressure and 3g inertia load (lifting - cask in the vertical orientation)

The weight of the resin and aluminum containers is modeled as an additional pressure on the bottom inner surface. The added pressure load is 9.31 psi per g. The effect of the outer shell dead weight is accounted for by using a 3g gravitational load in the longitudinal direction. The maximum stress intensity for this load case is 6.57 ksi.

Stress due to 25 psig internal pressure and 1 foot end drop (12g) (cask in the vertical orientation)

Page 2.10.3-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The weight of the resin and aluminum containers is modeled as an additional pressure on the bottom inner surface. The effect of the outer shell dead weight is accounted for by using a 12g gravitational load in the longitudinal direction using the same added pressure per g as used for the lifting case. The maximum shell stress intensity for this load case is 17.0 ksi.

Cask in the Horizontal Orientation

Stresses due to 25 psig internal pressure The stress due to 25 psig internal pressure is the same in both horizontal and vertical orientations (5.02 ksi).

Stress due to 25 psig internal pressure, 2g vertical, 5g lateral, and 10g longitudinal forward acceleration.

The vertical and lateral accelerations are combined such that g = (2.02 + 5.02)1/2 = 5.4g.

When the cask is in horizontal orientation, it is assumed that the weight of the outer shell, resin, and aluminum containers of the top 140° is supported by gamma shield, and the remaining weight of the outer shell, resin, and aluminum containers (220°) is uniformly distributed over the 152.5 in. length and over 180° arc.

For the loading due to 10g longitudinal forward acceleration, the weight of the resin and aluminum boxes is modeled as an additional pressure on the forward (lid side) inner surface of the outer shell.

The effect of the outer shell dead weight is accounted for by using a 10g gravitational load in the longitudinal direction. The maximum shell stress intensity for this load case is 17.64 ksi.

Stress due to 25 psig internal pressure, 2g vertical, 5g lateral accelerations, and 10g longitudinal backward acceleration.

The loading due to 25 psig internal pressure, 2g vertical, and 5g lateral and 10g longitudinal backward acceleration is the same as it is in the previous case.

The effect of the outer shell dead weight is accounted for by using a 10g gravitational load in the longitudinal direction. The maximum stress intensity for this load case is 17.54 ksi.

Stress due to 25 psig internal pressure and 1 foot side drop (16g)

When calculating the stress due to a 16g inertia load, it is assumed that the weight of the outer shell, resin, and aluminum containers of the top 140° is supported by the gamma shield, and the remaining weight of the outer shell, resin, and aluminum containers (220°) is uniformly distributed over the 152.5 in. length and over 180° arc.

The maximum stress intensity for this load case is 11.62 ksi.

Page 2.10.3-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Based on the above calculations the shell stress intensities are summarized in the following table:

2.10.3.6 Analysis Results Stress Intensities Loading (ksi) 25 psig Internal Pressure 5.02 25 psig + 3g Down 6.57 (Cask in Vertical Orientation) 25 psig + 12g Down 17.00 (Vertical End Drop) 25 psig + 2g vertical and 5g Lateral + 10g Longitudinal (Forward) 17.64 (Cask in Horizontal Orientation) 25 psig + 2g vertical and 5g Lateral + 10g Longitudinal (Backward) 17.54 (Cask in Horizontal Orientation) 25 psig + 16g Down 11.62 (Side Drop)

All the above calculated maximum stress intensities are less than the allowable stress of 27.0 ksi (1.5 Sm, SA-516 GR.55, at 250°F).

Page 2.10.3-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Weld stress intensities are also calculated at the locations noted in Figure 2.10.3-1.

These values are shown below.

Weld Location Max. Stress (Figure 2.10.3-1) Intensity (ksi) 25 psig Internal Pressure Location 1 2.38 Location 2 5.02 Location 3 4.78 Location 4 2.31 25 psig Internal Pressure + Lifting Load Location 1 7.34 Location 2 7.43 Location 3 3.93 Location 4 2.38 25 psig Internal Pressure + End Drop Loads Location 1 13.86 Location 2 11.21 Location 3 1.40 Location 4 13.87 25 psig Internal Pressure + Forward Acceleration Loads Location 1 13.39 Location 2 2.30 Location 3 11.03 Location 4 13.95 25 psig Internal Pressure + Backward Acceleration Loads Location 1 14.27 Location 2 11.02 Location 3 2.02 Location 4 13.11 25 psig Internal Pressure + Side Drop Loads Location 1 9.40 Location 2 8.93 Location 3 8.50 Location 4 9.19 The weld stress intensities are less than the allowable stress of 16.5 ksi (0.3Su, SA-516 GR.55, at 250° F).

Page 2.10.3-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.3.7 References

1. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 1989.

2. 10 CFR 71 Packaging and Transportation of Radioactive Material.

3. ANSYS Engineering Analysis System, Users Manual for ANSYS Release 8.0.

Page 2.10.3-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.3-1 Cask Outer Shell and Connection to Cask Body Page 2.10.3-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.3-2 Finite Element Model - Boundary Conditions Page 2.10.3-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Appendix 2.10.4 Fracture Toughness Evaluation of the TN-40 Cask TABLE OF CONTENTS 2.10.4.1 Introduction........................................................................ 2.10.4-1 2.10.4.2 Fracture Toughness Requirements of The Cask ............... 2.10.4-1 2.10.4.3 Fracture Toughness Evaluation of Cask Components and Welds ......................................................................... 2.10.4-3 2.10.4.4 Methodology ...................................................................... 2.10.4-3 2.10.4.5 Loadings ............................................................................ 2.10.4-3 2.10.4.6 Material Fracture Toughness ............................................. 2.10.4-3 2.10.4.7 Fracture Toughness Criteria .............................................. 2.10.4-5 2.10.4.8 Stress Intensity Factor Calculations .................................. 2.10.4-5 2.10.4.9 Conclusions ....................................................................... 2.10.4-5 2.10.4.10 NDE Inspection Plan ......................................................... 2.10.4-6 2.10.4.11 References ........................................................................ 2.10.4-7 LIST OF TABLES Table 2.10.4-1 Summary Stress Components - Normal Conditions of Transport ................................................................................ 2.10.4-8 Table 2.10.4-2 Summary Stress Components - Hypothetical Accident Conditions ............................................................................... 2.10.4-9 Table 2.10.4-3 Summary Stress Intensity Factors for Normal Condition of Transport Loadings ............................................................... 2.10.4-10 Table 2.10.4-4 Summary Stress Intensity Factors for Hypothetical Accident Condition Loadings ................................................ 2.10.4-11 Page 2.10.4-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 2.10.4-1 Critical Locations for Stress and Fracture Evaluation ........... 2.10.4-12 Figure 2.10.4-2 Charpy V-Notch Test Results for SA-266 Forging ................ 2.10.4-13 Page 2.10.4-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 FRACTURE TOUGHNESS EVALUATION OF THE TN-40 CASK 2.10.4.1 Introduction This appendix documents the fracture toughness evaluation of the TN-40 cask.

2.10.4.2 Fracture Toughness Requirements of The Cask The TN-40 cask material is a ferritic steel (penetration covers are stainless steel) and is therefore subject to fracture toughness requirements in order to assure ductile behavior at the lowest service temperature (LST) of -20 °F.

The inner shell and bottom inner plate are fabricated from SA-203 Gr. D or E plate material, 1.5 inches thick. The shell flange is 4.6 inches thick, fabricated from SA-350 Gr. LF3 forging material and the lid outer plate is 4.5 inches thick, fabricated from either SA-350 Gr. LF3 or SA-203 Gr. E material. The 1.5-inch lid closure bolts are fabricated from SA540 Grade B23/B24, Class1 material.

By interpolating between values provided in NUREG/CR-3826 [1] and NUREG/CR-1815

[2], the nil ductility transition temperatures (TNDT) of the containment boundary materials are:

Inner Shell and bottom inner plates (1.5 in.): -80 °F

Shell Flange (4.6 in.): -137 °F

Lid Outer Plate (4.5 in.): -125 °F The fracture toughness requirements of the lid closure bolts meet the criteria of ASME Code,Section III, Subsection NB (Para. NB-2333) [3]. Charpy v-notch testing is performed at -20 °F. The acceptance criterion is that the material exhibits at least 25 mils lateral expansion (Table NB-2333-1). All the lid closure bolt materials meet the NB-2300 criteria.

The 1.5 in. plate material which forms the inner shell and inner bottom plate meets the NUREG fracture arrest criteria.

Drop weight and Charpy test measurements of the shell flange and lid outer plate from 24 TN-40 casks are shown in the table below.

Page 2.10.4-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Measured Data for Shell Flange and Lid Outer Plate SHELL FLANGE LID OUTER PLATE Charpy Test Result Charpy Test Result Measured Test Ave. Ave. Lateral Measured Test Ave. Ave. Lateral Cask # TNDT Temp. Energy Expansion TNDT Temp. Energy Expansion

(°F) (°F) (ft-lbs) (mils) (°F) (°F) (ft-lbs) (mils) 1 -80 -20 107 73 -80 -20 88 60 2 -80 -20 107 73 -80 -20 84 75 3 -80 -20 107 73 -80 -20 97 67 4 -80 -20 118 79 -80 -20 95 60 5 -80 -20 102 67 -80 -20 101 73 6 -80 -20 102 67 -80 -20 68 51 7 -80 -20 112 77 -80 -20 62 51 8 -90 -30 107 88 -110 -50 110 79 9 -90 -30 107 88 -110 -50 110 79 10 -90 -30 107 88 -110 -50 110 79 11 -90 -30 107 88 -110 -50 110 79 12 -90 -30 107 88 -110 -50 110 79 13 -112 -52 150 92 -126 -67 180 93 14 -112 -52 150 92 -126 -67 180 93 15 -112 -52 150 92 -126 -67 180 93 16 -112 -52 150 92 -126 -67 180 93 17 -112 -52 150 92 -126 -67 180 93 18 -134 -83 143 63 -136 -83 106 60 19 -134 -83 143 63 -136 -83 106 60 20 -134 -83 143 63 -136 -83 106 60 21 -136 -83 126 61 -80 -20 259 96 22 -136 -83 126 61 -80 -20 259 96 23 -136 -83 126 61 -80 -20 259 96 24 -136 -83 126 61 -80 -20 259 96 From the measured data for the TN-40 cask containment boundary material described above, one can observe that the actual TNDT of some of the lid outer plate and shell flange material does not meet the NUREG fracture arrest criteria. Therefore, a fracture mechanics evaluation is performed for all the applicable cask components (including non-containment boundary components).

Page 2.10.4-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.4.3 Fracture Toughness Evaluation of Cask Components and Welds A fracture toughness evaluation of the TN-40 cask components and welds based on a service temperature of -20°F is performed. The evaluation includes the following:

Methodology

Loadings

Material fracture toughness

Fracture toughness criteria

Stress Intensity Factor calculations

Conclusions

NDE Inspection Plan 2.10.4.4 Methodology The allowable flaw sizes were determined using linear elastic fracture mechanics (LEFM) methodology from Section XI of ASME Code [4]. Flaws in the welds, if they occur, are welding defects, rather than initiated cracks. There is no active mechanism for crack initiation and growth at any of the weld locations since all the containment welds are volumetrically examined by RT and/or UT examination to assure no weld defects are present.

2.10.4.5 Loadings Figure 2.10.4-1 shows the selected locations on the cask numbered 1 through 10 for fracture toughness analysis. Stresses are linearized at these critical locations for maximum tensile membrane and bending stresses. Table 2.10.4-1 and Table 2.10.4-2 list the maximum membrane and bending stresses at these selected locations under NCT and HAC.

2.10.4.6 Material Fracture Toughness The shell flange is basically a forged cylinder, nominally 4.6 inches thick by 9 inches long, made from SA-350 Gr. LF-3 material. The welding of the flange to the shell may be performed using SAW, FCAW, or GTAW processes. The lid outer plate is either a forged disc or a plate, nominally 4.5 inches thick with an 82.75 inch diameter, made from either SA-350 Gr. LF3 or SA-203 Gr. E material.

The Charpy impact testing data for twenty four TN-40 casks are tabulated above. The tabulated data shows a dispersion in the absorbed energy values mainly due to the flange and lid material being supplied by different material suppliers.

The electrodes used in the shell flange and lid outer plate weldments have a high nickel content. The high alloy content of the electrodes and their typical usage in applications where good toughness is required indicate that the expected fracture toughness values for the weld filler material is as good or better than that of the base material.

Page 2.10.4-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The gamma shield shell is a forged cylinder, nominally 8 inches thick by 167 inches long. The bottom shield is nominally 8.75 inches thick and 91 inches in diameter.

These components are made from SA-266 Class 4, SA-516 Gr. 70 or SA-105 material.

Similarly, the 6 in. thick lid shield plate is made from SA-516 Gr. 70 or SA-105. The welding at the top flange and bottom plate may be performed using SAW, FCAW, or SMAW processes.

The lid and shell flange materials have better fracture toughness properties than the SA-266 used for the gamma shield shell. Therefore, the low fracture toughness material SA-266 forging can be considered bounding. It is thus conservative to use the fracture toughness properties of the SA-266 forging as the basis for qualifying the TN40 containment material.

Reference 5 is a very thorough review of correlations between a range of ferritic steel material strength levels and Charpy impact energies. Figure 2.10.4-2 (reproduced from Figure 4-5 of [5]) plots Charpy V-notch impact test results for a normalized SA-266 forging. The actual data points are shown along with a smoothed line that connects the average value at each test temperature. This data demonstrates that a lower bound Charpy impact value of 18 ft-lbs is appropriate for an exposure temperature of -20°F.

The various correlations between Kic and Kid given in Table 4-2 of [5] are compared at the 18 ft-lb level. Using the equation for yield strength for 36 to 50 ksi in transition in Table 4-2 of [5], the Charpy impact measurement may be transformed into a fracture toughness value:

Kid = 5E(Cv)1/2 = 50,289 psi-(in)1/2 = 50 ksi-(in.)1/2 Where Kid = Dynamic Fracture Toughness (based on crack arrest), psi -(in)1/2 E = Modulus of Elasticity, 28.1 106 psi (conservatively use 300°F)

Cv = Charpy Impact Measurement, 18 ft-lbs For conservatism, the above calculated Kid was reduced to 47 ksi-(in)1/2 for fracture toughness evaluations of the TN-40 Cask components (containment and non-containment boundary) and welds.

SAW, FCAW and SMAW electrodes used in the gamma shield weldments are alloyed to provide good low temperature properties. Examples are AWS Class E71T-1 and F7P6-EH14 materials. Although Charpy testing is not a requirement for the gamma shield shell material, testing has been conducted by TN-40 fabricators to demonstrate the toughness of the forged material and the associated welds. The results show that all regions of the SA-516-70 weldments have fracture toughness values well in excess of the 18 ft-lbs specified. Thus, use of the fracture properties from the wrought material for locations at or near the weld joints is conservative.

Page 2.10.4-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.4.7 Fracture Toughness Criteria Using the rule of Section XI, IWB-3613 [4], the limiting fracture toughness values are reduced by a factor of 10 for the NCT and 2 for the HAC, to define the limiting allowable Kallowable. That is, Kallowable Kia/(10) = 47/(10) = 14.86 ksi-in for normal conditions Kallowable Kic /(2) = 47/(2) = 33.23 ksi-in for accident conditions Where:

Kia = the available fracture toughness based on crack arrest for the corresponding crack tip temperature Kic = the available fracture toughness based on crack initiation for the corresponding crack tip temperature Because of the dynamic loading (1-foot and 30-foot drops), it is appropriate to use the Kid value (47 ksi-in1/2) calculated above for Kia and Kic for the following normal and accident condition fracture toughness evaluations.

2.10.4.8 Stress Intensity Factor Calculations The total applied stress intensity KI (applied) is determined from the membrane and bending stresses. For purpose of analysis, the postulated surface flaws are oriented in both the axial and circumferential directions. The surface crack depth is assumed as 15% of component thickness. However, the maximum crack depth is limited to 1/2 inch.

The crack length is assumed to be 10 times the crack depth. The assumed crack sizes are such that they can be readily spotted by visual examination. Compared to surface cracks, same size subsurface cracks are less critical. The results of the applied stress intensity KI calculations for normal and accident conditions are shown in Table 2.10.4-3 and Table 2.10.4-4, respectively.

2.10.4.9 Conclusions Based on the results of fracture analysis of TN-40 cask components and welds with the postulated surface crack sizes, it is concluded that there is no potential of fracture failure due to NCT and HAC transport loadings. The postulated surface flaw sizes are such that they can be readily detected by a visual inspection.

Page 2.10.4-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Note that the gamma shield shell is not part of the containment boundary. Cracks postulated in the gamma shield shell will not propagate into the containment boundary due to the geometry of the cask. If the gamma shield shell were to fracture along the length or around the circumference or around the weld between the gamma shield shell and top flange, there is no credible mechanism that would result in the gamma shielding separating from the containment vessel. The top shield plate is welded to the lid and is captured by the containment vessel. Therefore, if the weld were to completely fail, the shield plate would still remain inside the containment boundary and would not lose its shielding capability. Therefore, even if a fracture were to occur in the gamma shield shell or the weld between the gamma shield and top flange or top shield plate or weld between top shield plate and lid, there would be no safety significance, since containment would be maintained, and shielding would remain in place. The one exception is in the region of the weld of the gamma shield shell to the bottom plate. In this region, if the weld were to completely fail, the bottom plate could become detached and have an impact on the shielding capability of the cask. However, the bottom trunnions are independently welded to the gamma shield shell and the bottom shield plate and these additional attachment points (welds) would resist detachment of the bottom shield plate from the cask.

2.10.4.10 NDE Inspection Plan The results of the fracture toughness analysis show that the flaws in the gamma shield shell and top and bottom shield plates which would result in unstable crack growth or brittle fracture are larger than those generally observed in forged steel and plate components. No special examination requirements on the gamma shield shell, top and bottom shield plates are, therefore, required.

The flaw sizes in the welds that could result in brittle fracture at -20F will be detected by NDE methods and repaired.

The liquid penetrant or magnetic particle method will be in accordance with Section V, Article 6 of ASME Code [4].

Page 2.10.4-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.4.11 References

1. NUREG/CR-3826, Recommendations for Protecting against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, April 1984.

2. NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers up to Four Inches Thick.

3. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code Section III, Subsection NB, 1989.

4. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code Section V and Section XI, 1989.

5. SIR-98-110, Rev. 0, Allowable Flaw Evaluation of Transnuclear TN-32 Cask, Structural Integrity Associates, Inc. 1998.

Page 2.10.4-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.4-1 Summary Stress Components - Normal Conditions of Transport Cask Location Membrane Stress (ksi) Bending Stress (ksi)

Max. Stress from SX SY SZ SX SY SZ Type*

Figure 2.10.4-1 (Rad.) (Tang.) (Axial) (Rad.) (Tang.) (Axial)

SY (N11) 0.03 1.74 -0.61 0.08 0.10 0.35

1. Weld SZ(N11) 0.01 0.73 -0.53 0.24 0.04 0.31 SY(N11) -1.54 0.37 -0.50 0.87 0.42 0.50

2. Weld SZ(N11) -1.51 0.23 -0.33 0.86 0.40 0.51 SY(N11) 1.23 1.51 0.80 0.24 0.50 1.11

3. Weld SZ(N11) 3.26 0.71 1.36 0.77 0.68 1.37 SX(N11) -0.50 -1.03 -0.14 3.59 1.03 1.52

4. Bottom Shield SY(N11) 0.25 1.54 -0.21 0.87 1.41 0.01 SZ(N11) -2.12 -1.32 1.55 2.13 1.45 2.24

5. Gam. Shield SY(N10) -0.40 3.21 1.90 0.33 2.20 2.20 Shell-Mid SZ(N10) -0.41 3.21 1.89 0.35 2.15 2.23

6. Gam. Shield SY(N10) -0.40 3.45 1.55 0.34 2.06 2.62 Shell-End SZ(N10) -0.44 -1.32 1.39 2.62 1.98 4.29 SX,SY(N9) 0.23 0.17 0.13 3.15 1.46 1.55

7. Flange SZ(N9) -0.06 0.51 2.54 2.21 1.32 2.57 SX(N11) 0.38 -0.01 -0.41 1.42 1.24 0.10

8. Shield Plate SY(N11) 0.27 0.80 -0.37 1.12 1.64 0.14 SZ(N11) -0.20 -0.99 0.34 0.27 0.22 0.12 SX(N11) 0.26 -0.04 0.67 2.03 0.46 0.78

9. Lid Outer Plate SY(N11) 0.44 1.06 2.16 0.30 1.36 1.46 SZ(N11) 0.62 1.37 2.98 1.28 0.32 0.48

10. Inner Shell & SX(N7) -0.64 -3.29 -4.00 2.91 1.23 0.25 Bottom Inner SY(N7) -0.66 -1.0 3.32 2.59 1.27 0.60 Plate SZ(N7) -0.71 -1.51 3.71 2.20 1.06 0.53

SY - Tangential Stress SZ - Axial Stress Page 2.10.4-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.4-2 Summary Stress Components - Hypothetical Accident Conditions Cask Location Max. Membrane Stress (ksi) Bending Stress (ksi) from Stress SX SY SZ SX SY SZ Figure 2.10.4-1 Type* (Rad.) (Tang.) (Axial) (Rad.) (Tang.) (Axial)

1. Weld SY, SZ (A13) 0.14 3.99 -0.71 0.98 0.56 0.78 SY(A12) -1.66 3.34 -1.89 2.21 0.93 0.90

2. Weld SZ(A12) -0.21 -0.35 -0.45 0.07 0.10 0.07 SY(A12) -10.52 0.51 -3.80 2.59 1.04 1.86

3. Weld SZ(A12) 0.40 -1.52 -0.18 1.17 0.41 0.04 SX(A5) -5.20 -7.31 0.34 8.92 3.68 2.08

4. Bottom Shield SY(A5) -1.32 2.27 -0.94 3.20 3.17 1.50 SZ(A5) -7.01 -7.71 2.74 5.57 4.00 3.67

5. Gam. Shield SY(A11) -0.43 4.32 2.62 0.41 1.13 1.22 Shell-Mid SZ(A11) -0.43 4.32 2.62 0.41 1.13 1.22

6. Gam. Shield SY(A11) 1.12 4.53 0.01 2.16 3.81 5.62 Shell-End SZ(A11) 1.06 0.10 1.19 1.17 1.02 3.79 SX,SY(A11) -0.03 4.67 -0.35 0.11 1.10 0.27

7. Flange SZ(A11) 0.06 6.46 3.95 0.28 1.47 3.08 SX,SY(A12) -4.77 -3.98 -0.76 16.55 11.82 0.38

8. Shield Plate SZ(A12) -2.74 -1.52 8.99 8.26 6.40 10.92 SX(A12) 0.26 6.55 -3.35 2.14 0.15 3.34

9. Lid Outer Plate SY(A12) -3.74 4.45 -2.70 11.50 5.87 2.79 SZ(A12) 1.01 5.45 2.93 0.07 3.14 0.22

10. Inner Shell & SX(A8) 0.78 -7.65 -7.35 6.87 1.82 1.19 Bottom Inner Plate SY, SZ (A8) 1.98 5.40 7.16 4.72 2.13 0.40

SY - Tangential Stress SZ - Axial Stress Page 2.10.4-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.4-3 Summary Stress Intensity Factors for Normal Condition of Transport Loadings Allow.

Crack Crack Stress Thick., Max. Critical Stress Factors depth, length, Intens.

Component t Stress Crack Intens. of a l Factor (in) Type (1) Direction (1) Factor Safety (in) (in) (ksi-in)

(ksi-in)

1. Weld 1.25 0.1875 1.875 SY Axial 1.52 14.86 9.77

2. Weld 0.5 0.075 0.75 SY Axial 0.37 14.86 >10

3. Weld 0.75 0.1125 1.125 SZ Hoop 1.59 14.86 9.35

4. Bottom Shield 8.75 0.500 5.00 SZ Hoop 4.73 14.86 3.15

5. Gam. Shield Shell-8.0 0.500 5.00 SY Axial 6.89 14.86 2.16 Mid

6. Gam. Shield Shell-8.0 0.500 5.00 SZ Hoop 7.18 14.86 2.07 End

7. Flange 4.6 0.500 5.00 SZ Hoop 6.32 14.86 2.36

8. Shield Plate 6.0 0.500 5.00 SY Axial 2.92 14.86 5.10

9. Lid Outer Plate 4.5 0.500 5.00 SZ Hoop 5.65 14.86 2.64

10. Inner Shell & Bottom 1.5 0.225 2.25 SZ Hoop 3.86 14.86 3.86 Inner Plate (1) SY - Hoop Stress - Results in Axial crack SZ - Axial Stress - Results in Hoop crack Page 2.10.4-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.4-4 Summary Stress Intensity Factors for Hypothetical Accident Condition Loadings Allow.

Crack Crack Stress Thick., Max. Critical Stress Factors depth, length, Intens.

Component t Stress Crack Intens. of a l Factor (in) Type (1) Direction (1) Factor Safety (in) (in) (ksi-in)

(ksi-in)

1. Weld 1.25 0.1875 1.875 SY Axial 3.74 33.23 8.90

2. Weld 0.5 0.075 0.75 SY Axial 2.17 33.23 >10

3. Weld 0.75 0.1125 1.125 SZ Hoop 0.97 33.23 >10

4. Bottom Shield 8.75 0.500 5.00 SX Tang. 10.71 33.23 3.10

5. Gam. Shield Shell-8.0 0.500 5.00 SY Axial 7.08 33.23 4.69 Mid

6. Gam. Shield Shell-8.0 0.500 5.00 SY Axial 9.85 33.23 3.37 End

7. Flange 4.6 0.500 5.00 SY Axial 10.36 33.23 3.20

8. Shield Plate 6.0 0.500 5.00 SZ Hoop 25.04 33.23 1.33

9. Lid Outer Plate 4.5 0.500 5.00 SY Axial 15.65 33.23 2.12

10. Inner Shell &

1.5 0.225 2.25 SZ Hoop 7.51 33.23 4.42 Bottom Inner Plate (1) SY: Hoop Stress - Results in Axial crack SZ: Axial Stress - Results in Hoop crack Sx: Radial Stress - Results in Tangential Crack Page 2.10.4-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.4-1 Critical Locations for Stress and Fracture Evaluation Page 2.10.4-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.4-2 Charpy V-Notch Test Results for SA-266 Forging Page 2.10.4-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.5 STRUCTURAL ANALYSIS OF THE TN-40 BASKET TABLE OF CONTENTS 2.10.5.1 Introduction........................................................................ 2.10.5-1 2.10.5.2 TN-40 Fuel Basket Stress Analysis ................................... 2.10.5-3 2.10.5.3 TN-40 Fuel Basket Buckling Analysis .............................. 2.10.5-11 2.10.5.4 Fusion Welds ................................................................... 2.10.5-14 2.10.5.5 Sensitivity Analysis .......................................................... 2.10.5-15 2.10.5.6 Summary ......................................................................... 2.10.5-19 2.10.5.7 References ...................................................................... 2.10.5-24 LIST OF TABLES Table 2.10.5-1 Material Properties for TN-40 Fuel Basket ............................ 2.10.5-25 Table 2.10.5-2 Basket Structural Allowable Stresses, NCT .......................... 2.10.5-26 Table 2.10.5-3 Basket Structural Allowable Stresses, HAC .......................... 2.10.5-26 Table 2.10.5-4 NCT 0° Side Drop, Basket Stress Analysis Results .............. 2.10.5-27 Table 2.10.5-5 NCT 45° Side Drop, Basket Stress Analysis Results ............ 2.10.5-28 Table 2.10.5-6 NCT 90° Side Drop, Basket Stress Analysis Results ............ 2.10.5-29 Table 2.10.5-7 HAC 0° Side Drop, Basket Stress Analysis Results .............. 2.10.5-30 Table 2.10.5-8 HAC 45° Side Drop, Basket Stress Analysis Results ............ 2.10.5-31 Table 2.10.5-9 HAC 90° Side Drop, Basket Stress Analysis Results ............ 2.10.5-32 Table 2.10.5-10 DELETED ............................................................................. 2.10.5-32 Table 2.10.5-11 Fuel Basket Buckling Analysis Results ................................. 2.10.5-33 Page 2.10.5-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 2.10.5-1 Typical TN-40 Basket Fuel Compartment Plate .................... 2.10.5-34 Figure 2.10.5-2 TN-40 Basket Finite Element Model Including Drop Orientations .......................................................................... 2.10.5-35 Figure 2.10.5-3 Basket Finite Element Model Displacement Constraints -

0° Side Drop ......................................................................... 2.10.5-36 Figure 2.10.5-4 Basket Finite Element Model Displacement Constraints -

30° Side Drop ....................................................................... 2.10.5-37 Figure 2.10.5-5 Basket Finite Element Model Displacement Constraints -

45° Side Drop ....................................................................... 2.10.5-38 Figure 2.10.5-6 Basket Finite Element Model Displacement Constraints -

60° Side Drop ....................................................................... 2.10.5-39 Figure 2.10.5-7 Basket Finite Element Model Displacement Constraints -

90° Side Drop ....................................................................... 2.10.5-40 Figure 2.10.5-8 Basket Finite Element Model Temperature Boundary Condition............................................................................... 2.10.5-41 Figure 2.10.5-9 Basket Finite Element Model Applied Pressures - 0° Drop, NCT ...................................................................................... 2.10.5-42 Figure 2.10.5-10 Basket Finite Element Model Applied Pressures - 45° Drop, NCT............................................................................. 2.10.5-43 Figure 2.10.5-11 Basket Finite Element Model Applied Pressures - 90° Drop, NCT............................................................................. 2.10.5-44 Figure 2.10.5-12 NCT 45° Side Drop - S.S. Plates - Membrane plus Bending Stress Intensity ..................................................................... 2.10.5-45 Figure 2.10.5-13 NCT 45° Side Drop - Al. Plates - Membrane plus Bending Stress Intensity ..................................................................... 2.10.5-46 Figure 2.10.5-14 NCT 45° Side Drop - Al. Periphery Plates - Membrane plus Bending Stress Intensity................................................ 2.10.5-47 Figure 2.10.5-15 NCT 45° Side Drop - Al. Outer Plates - Membrane plus Bending Stress Intensity ....................................................... 2.10.5-48 Figure 2.10.5-16 Thermal Stress Analysis Finite Element Model -

Configuration 1 ..................................................................... 2.10.5-49 Figure 2.10.5-17 Thermal Stress Analysis Finite Element Model -

Configuration 2 ..................................................................... 2.10.5-50 Figure 2.10.5-18 Thermal Stress Analysis - Configuration 1 - Maximum Aluminum Stress Intensity .................................................... 2.10.5-51 Figure 2.10.5-19 Thermal Stress Analysis - Configuration 1 - Maximum Stainless Steel Stress Intensity............................................. 2.10.5-52 Page 2.10.5-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-20 Thermal Stress Analysis - Configuration 2 - Maximum Aluminum Stress Intensity .................................................... 2.10.5-53 Figure 2.10.5-21 Thermal Stress Analysis - Configuration 2 - Maximum Stainless Steel Stress Intensity............................................. 2.10.5-54 Figure 2.10.5-22 Fuel Basket Buckling Analysis Finite Element Model Node Couplings .............................................................................. 2.10.5-55 Figure 2.10.5-23 Basket Buckling Analysis Loading Boundary Conditions -

0° Side Drop ......................................................................... 2.10.5-56 Figure 2.10.5-24 Basket Buckling Analysis Loading Boundary Conditions -

30° Side Drop ....................................................................... 2.10.5-57 Figure 2.10.5-25 Basket Buckling Analysis Loading Boundary Conditions -

45° Side Drop ....................................................................... 2.10.5-58 Figure 2.10.5-26 Basket Buckling Analysis Loading Boundary Conditions -

60° Side Drop ....................................................................... 2.10.5-59 Figure 2.10.5-27 Basket Buckling Analysis Loading Boundary Conditions -

90° Side Drop ....................................................................... 2.10.5-60 Figure 2.10.5-28 Fuel Basket Deformation at Buckling Load - 0° Side Drop ... 2.10.5-61 Figure 2.10.5-29 Fuel Basket Deformation at Buckling Load - 30° Side Drop . 2.10.5-62 Figure 2.10.5-30 Fuel Basket Deformation at Buckling Load - 45° Side Drop . 2.10.5-63 Figure 2.10.5-31 Fuel Basket Deformation at Buckling Load - 60° Side Drop . 2.10.5-64 Figure 2.10.5-32 Fuel Basket Deformation at Buckling Load - 90° Side Drop . 2.10.5-65 Figure 2.10.5-33 Fuel Compartment Interface Elements used in Sensitivity Study .................................................................................... 2.10.5-66 Figure 2.10.5-34 Basket Deformation at Buckling Load for 0 Degree Side Drop ...................................................................................... 2.10.5-67 Figure 2.10.5-35 Initial Imperfections Imposed ................................................ 2.10.5-68 Figure 2.10.5-36 Basket Deformation at Buckling Load for 0 Degree Side Drop with Initial Imperfections .............................................. 2.10.5-69 Figure 2.10.5-37 Basket Finite Element Model to Calculate the Maximum Compressive Load ................................................................ 2.10.5-70 Page 2.10.5-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 STRUCTURAL ANALYSIS OF THE TN-40 BASKET 2.10.5.1 Introduction This appendix presents the structural analysis of the TN-40 fuel support basket. The basket is a welded assembly of stainless steel boxes and is designed to accommodate 40 PWR fuel assemblies. The fuel compartment stainless steel box sections are attached together locally by cylindrical stainless steel plugs that pass through the aluminum and Boral plates and are fusion welded to both adjacent box sections. The basket contains 40 compartments for proper spacing and support of the fuel assemblies.

The basket structure is open at each end and therefore, longitudinal fuel assembly loads are applied directly to the cask body and not to the fuel basket structure. The fuel assemblies are laterally supported by the stainless steel structural tubes, and the basket is laterally supported by the cask inner shell.

The deformations and stresses induced in the basket structure due to the applied lateral loads are determined using the ANSYS computer program [1]. The most severe loadings for which the basket is evaluated are the 30 foot Hypothetical Accident Condition (HAC) side drop and end drop accidents. The basket is also evaluated for 1 foot side drop and end drop loads under the Normal Conditions of Transport (NCT).

The g-loads and drop orientations used for the basket structural analysis are described in Appendix 2.10.8, Appendix 2.10.9, and Section 2.7.1 of Chapter 2. The dynamic load factor is calculated in Appendix 2.10.6. The inertial loads of the fuel assemblies are applied as equivalent pressures on the stainless steel box interior surfaces. Quasi-static stress analyses are performed with applied loads in equilibrium with the reactions at the periphery of the basket. The calculated stresses in the basket structure are compared with the stress limits to demonstrate that the established design criteria are met.

A summary of the accident analyses performed is presented in Section 2.10.5.6.

2.10.5.1.1 TN-40 Fuel Basket Geometry The details of the TN-40 basket are shown on TN Drawing Nos. 10421-71-8 and -9 in Chapter 1, Appendix 1.4.1. As described above, the basket structure consists of an assembly of stainless steel boxes or cells joined by fusion welded steel plugs and separated by aluminum and neutron poison material (Boral sheets). The stainless, aluminum and Boral wall between fuel compartments is effectively a sandwich panel.

The 304 stainless steel members are the primary structural components. The aluminum provides the heat conduction path from the fuel assemblies to the cask inner shell, and the neutron poison material provides the necessary criticality control.

Page 2.10.5-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 A representative basket wall panel between fuel compartments is shown in Figure 2.10.5-1. The panel plates are welded together at discrete locations (2 attachments every 8 inches) along their length. The adjacent fuel compartment stainless steel walls are fusion welded to cylindrical plugs that pass through holes in the Boral and aluminum plates. This method of construction forms a very strong honeycomb-like structure of boxes. The nominal open dimension of each fuel compartment cell or box is 8.05 in. x 8.05 in. which provides a minimum of 1/8 in. clearance around the fuel assemblies. The pitch of the cells is approximately 8.85 in. The overall basket length (160 in.) is less than the cask cavity length to allow for loading the fuel assemblies, thermal expansion and tolerance stackup.

Several of the aluminum conductor plates are continuous across the diameter of the basket to provide uninterrupted heat conduction paths. Other shorter plates are provided between and perpendicular to these continuous plates. Some of the aluminum plates are as short as one cell width.

Structural rails oriented parallel to the axis of the cask are attached to the inner cavity wall of the cask body to establish and maintain basket orientation, to prevent twisting of the basket assembly, and to support the edges of those plates adjacent to the rails which would otherwise be free to slide tangentially around the cask cavity wall under lateral inertial loadings.

2.10.5.1.2 Fuel Basket Analysis Overview The fuel basket is evaluated for NCT and HAC impact and thermal loads. The basket stress analysis is performed using a finite element method for the side drop and thermal load cases and analytical calculations for the end drop load cases. Buckling of the basket plates when subjected to lateral impact loads is evaluated using a nonlinear finite element buckling analysis. Stress and buckling analyses are provided in Sections 2.10.5.2 and 2.10.5.3 respectively. Fusion weld testing program is provided in Section 2.10.5.4. Sensitivity study of the buckling analysis where nodal couplings were replaced with contact elements and the effect of initial imperfection of the fuel boxes are provided in Section 2.10.5.5.

2.10.5.1.3 Weight The total weight of the TN-40 basket is 14,693 lb., and the total weight of all 40 fuel assemblies is 52,000 lb. A value of 1,300 lb. is assumed for the weight of each fuel assembly. Under lateral inertial loading each assembly is assumed to be uniformly supported across the width and along the length of the tube wall. The inertia of the basket structure (weight of the basket x g-load) is also included in the analysis.

2.10.5.1.4 Temperature Thermal analyses are performed to obtain the temperature distributions in the basket for various conditions. These analyses are presented in Chapter 3. The model temperature distribution is shown in Figure 2.10.5-8. Thermal stresses induced in the fuel basket by the applied temperature distributions are evaluated in Section 2.10.5.2.4.

Page 2.10.5-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.5.2 TN-40 Fuel Basket Stress Analysis 2.10.5.2.1 Approach Bounding inertial loads of 20g and 75g are applied for the NCT and HAC transport cask free drop cases respectively. 0°, 45° and 90° azimuth orientations are analyzed in order to bound all possible drop orientations.

Nonlinear analyses with bilinear material properties and small deflections were performed in ANSYS [1] for the critical azimuth side drop orientations. The membrane and membrane plus bending stresses were compared against Sm and 1.5 Sm stress criteria values [2] for the NCT cask drop. The membrane and membrane plus bending stresses were compared against 0.7 Su and 0.9 Su elastic-plastic analysis stress criteria values [2] for the HAC cask drop.

The TN-40 transport cask geometry is described in Section 2.10.5.1.1 and depicted in detail in the design drawings provided in Chapter 1, Appendix 1.4.1 (Drawings 10421-71-8 and -9). Nominal dimensions are used in the analyses that follow.

Side drop impact analyses using finite element methods are provided in Section 2.10.5.2.2 and the analytical analysis for the end drop impact is provided in Section 2.10.5.2.3. The thermal stress analysis of the fuel basket is provided in Section 2.10.5.2.4.

2.10.5.2.2 Basket Finite Element Analysis for Side Impact Loads A) Finite Element Model Description A three-dimensional finite element model of the fuel basket is constructed using shell elements. The overall finite element model of the fuel basket is shown in Figure 2.10.5-2. The fuel tubes, aluminum structural plates, aluminum outer plates and periphery plates are included in the model. For conservatism, the strength of Boral plates in the basket is neglected by excluding these from the finite element model.

However, their weight is accounted for by increasing the structural aluminum plate material densities.

Because of the large number of plates in the basket and large size of the basket, certain modeling approximations were necessary. In view of continuous support of fuel compartment tubes by the peripheral rails along the entire basket length during a side drop, only an 8.0 inch long slice of the basket and rail is modeled. At the two cut faces of the model, symmetry boundary conditions are applied (UY = ROTX = ROTZ = 0).

The displacement constraints for the 0°, 45°, and 90° side drop angles are shown in Figure 2.10.5-3, Figure 2.10.5-5, and Figure 2.10.5-7 respectively. For clarity, symmetry displacement constraints are not shown.

Page 2.10.5-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The nodes between the steel tubes including the intermediate aluminum plates are coupled together in the out-of-plane direction so that they will bend in unison under surface pressure or other lateral loading to simulate through the thickness support provided by Boral plates. The aluminum plates are coupled together at their intersection. The fusion welds, connecting the fuel compartments and plates, are modeled with short pipe elements connected at each end to adjacent fuel compartment boxes in all directions. The nodal couplings are shown in Figure 2.10.5-22. Furthermore, a sensitivity study on the effect of the nodal couplings was performed and is provided in Section 2.10.5.5.

B) Material Properties and Design Criteria The stainless steel boxes are constructed from SA-240, Gr. 304 stainless steel. The aluminum plates, outer plates and basket periphery plates are constructed from SB-209, 6061-T651 aluminum alloy. A bilinear stress-strain curve for SA-240 Type 304 stainless steel and SB-209 Type 6061-T651 aluminum alloy (EP/E = 0.05) is used for the basket plates.

Data from a stress-strain curve for SA-240 Type 304 was taken from NUREG/CR 0481

[5]. This data used as a basis for the 5% strain-hardening rate shown in the bilinear stress-strain curve used in the basket analysis. Using a 5% strain-hardening rate which is greater than those resulting from this data is conservative because it will result in higher stresses.

For the aluminum SB-209 Type 6061-T651, reference [6] gives elongations of 17 - 70%

with associated strain hardening rates less than 5%. In the case of aluminum, also, using a 5% strain-hardening rate is conservative because it will yield higher stresses.

Note that all Pm limits (0.7Su) are below Sy. Therefore the strain-hardening rate has no effect on Pm stresses.

Table 2.10.5-1 lists the material properties used in all analyses of the TN-40 fuel basket.

Table 2.10.5-2 and Table 2.10.5-3 summarize the stress criteria for the NCT and HAC events respectively.

C) Side Drop Loading Conditions The basket structure is analyzed for 0°, 45° and 90° azimuth side drops. Due to the basket structure symmetry, these orientations of side drops are assumed to envelop all other possible drop orientations.

A fuel assembly weight of 1,300 lb. is used in the analysis. A uniform fuel weight distribution is assumed over 144 inches, which is the active fuel length. An 8.0 inch sector of the basket assembly is modeled. The weight of the Boral plates is accounted for by increasing the density of the aluminum plates. The Boral plates stiffness is conservatively neglected in the analysis.

Page 2.10.5-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Temperatures at the cross section where the maximum temperatures occur in the basket are used, which are taken from the normal transfer condition (100°F) thermal analysis presented in Chapter 3. Figure 2.10.5-8 shows the temperature contour used in all analyses.

The load resulting from the fuel assembly weight is applied as pressure on the fuel compartment plates of the basket. For the 0° orientation, the pressure acts only on the horizontal plates, and for the 90° orientation, the pressure acts only on the vertical plates. For the 45° orientation, the pressure was divided into components that act on both horizontal and vertical plates of the basket. The pressures for all orientations are calculated below for 20g and 75g accelerations.

Page 2.10.5-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 0° and 90° Drop Orientations Pressure for 1g, p = Fuel assembly weight / (Panel span x Panel length)

= 1300 lb. / (8.14 in. x144 in.) = 1.109 psi Pressure for 20g = 20 x 1.109 = 22 psi Pressure for 75g = 75 x 1.109 = 83 psi 45° Orientation Pressure for 1g = p cos 45°

= 1.109 x 0.7071 = 0.7842 Pressure for 20g = 22 x 0.7071 = 16 psi Pressure for 75g = 83 x 0.7071 = 59 psi The load distributions for the 0°, 45° and 90° analyses for the NCT drops are shown in Figure 2.10.5-9 to Figure 2.10.5-11, respectively. The load distribution for the 0°, 45° and 90° HAC drop analyses are similar to those shown in Figure 2.10.5-9 to Figure 2.10.5-11, except that the applied pressures are scaled up to 75g accordingly.

The acceleration applied in each run are as follows.

Orientation Inertial Load (g) ax (g) ay (g) az (g) 20 20 0 0 0° 75 75 0 0 20 14.14 0 -14.14 45° 75 53.03 0 -53.03 20 0 0 -20 90° 75 0 0 -75 D) Side Drop Analysis and Results NCT Side Drop Analysis and Results Nonlinear analyses with bilinear material properties and small deflections were performed using ANSYS [1] for the 0°, 45°, and 90° drop orientations. Loads corresponding to 20g were applied in all analyses. It was confirmed that no area in the model was plastically deformed; hence all the analyses were linear elastic analyses.

Page 2.10.5-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The nodal stress intensity distribution in the stainless steel boxes, aluminum plates, aluminum periphery plates and aluminum outer plates are computed by ANSYS. The membrane plus bending stress intensity distributions for the NCT 45° azimuth drops are shown in Figure 2.10.5-12 through Figure 2.10.5-15 as representative sample of the resulting stresses. The shell middle surface nodal stress intensity is the membrane stress intensity and the top or bottom surface stress intensity is the membrane plus bending stress intensity. Allowable stress / stress at each node for membrane and membrane plus bending stresses at the temperature for that node are calculated. The location where allowable stress / stress is at a minimum (the location with minimum factor of safety) is determined. The results are summarized in Table 2.10.5-4 through Table 2.10.5-6.

The limiting stress (Pm + Pb + Q) in the stainless steel plates for normal conditions, 10.15 ksi (maximum thermal stress is 1.45 ksi from Section 2.10.5.2.4), is also lower then the criteria 59.6 ksi (3 Sm at 310° F). The limiting stress (Pm + Pb + Q) in the aluminum plates for normal conditions, 5.01 ksi (maximum thermal stress is 0.84 ksi from Section 2.10.5.2.4), is also lower then the criteria, 10.0 ksi (3 Sm at 440° F).

HAC Side Drop Analysis and Results Nonlinear analyses with bilinear material properties and small deflection were performed in ANSYS [1] for the 0°, 45°, and 90° drop orientations. Loads corresponding to 75g were applied in all analyses. The nodal stress intensity distributions throughout the fuel basket are qualitatively similar to those shown in Figure 2.10.5-12 through Figure 2.10.5-15, except with higher stress amplitudes.

Allowable stress / stress at each node for membrane and membrane plus bending stresses at the temperature for that node are calculated. The location where allowable stress / stress is at a minimum (the location with minimum factor of safety) is determined. The results are summarized in Table 2.10.5-7 through Table 2.10.5-9.

Maximum Relative deflection of the Basket The maximum relative deflections in the critical fuel compartment during the 75g side drops of 0°, 45°, and 90° orientations are summarized in the following table.

Page 2.10.5-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 x 1 z

3 4 2

Maximum Relative Deflection of the Basket Total Deflection at Locations (in) Relative Deflection Orientation 1 (ux) 2 (ux) 3 (uz) 4 (uz) 1-2 (ux) 3-4 (uz) 0 degree -0.051 -0.028 -0.0001 0.0009 -0.023 -0.0008 45 degree -0.025 -0.028 0.037 0.022 0.003 0.015 90 degree -0.0014 -0.0007 0.052 0.029 -0.0007 0.023 It is seen that the maximum total relative deflection in the critical box is in the order of 0.023 in. It should be noted that this is the deflection at 75g; permanent deformation will be significantly less.

2.10.5.2.3 Fuel Basket End Drop Analysis During an end drop, the fuel assemblies and fuel compartments are forced against the bottom or top of the TN-40 cask. It is important to note that, for any vertical or near vertical loading, the fuel assemblies react directly against the bottom or top end of the cask and not through the basket structure as in lateral loading. It is the weight of the basket that causes axial compressive stress during an end drop. Axial compressive stresses are conservatively computed first by assuming that all of the basket weight is reacted by the compartment tubes during an end drop and second, that all of the basket weight is reacted by the aluminum plates. A conservative basket weight of 15.0 kips (actual weight is 14.693 kips) is used in the end drop stress calculations.

Stainless Steel Basket Components Assuming that all of the weight is supported by the stainless steel fuel compartments we

have, Area of Steel Baskets = 40 x [(w + 2t)2 - w2]

= 40 x [(8.05 + 2 x 0.09)2 - 8.052] = 117.2 in2 Stress in steel baskets at 1g loading = P/A

= 15,000 / 117.2 = 128.0 psi Therefore, the stress generated in the stainless fuel compartments is summarized as follows.

Page 2.10.5-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Allowable Stress (Pm)

Axial Stress (ksi) at 450°F (ksi)

NCT (20g) 2.56 18.1 HAC (75g) 9.60 44.8 The maximum stress (Pm + Pb + Q) for normal condition, 4.01 ksi (maximum thermal stress is 1.45 ksi from Section 2.10.5.2.4), is also lower than the criteria 54.3 ksi (3 Sm).

Page 2.10.5-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Aluminum Basket Plates The weight is assumed to be totally supported by the aluminum basket plates.

Area of aluminum basket plate - 340.9 in.2 Stress in Aluminum Basket Plates at 1g loading = P/A

= 15000 / 340.9 = 44.0 psi Therefore, the stress generated in the aluminum basket plates is summarized as follows.

Allowable Stress (Pm)

Axial Stress (ksi) at 450°F (ksi)

NCT (20g) 0.88 3.1 HAC (75g) 3.30 8.6 The maximum stress (Pm + Pb + Q) for normal condition, 1.72 ksi (maximum thermal stress is 0.84 ksi from Section 2.10.5.2.4), is also lower than the criteria 9.2 ksi (3 Sm).

2.10.5.2.4 Fuel Basket Thermal Stress Analysis An elastic ANSYS [1] finite element analysis was conducted on the basket to evaluate the thermal stresses in the stainless steel and aluminum plates. Two different areas in the basket were analyzed, the first configuration included two 0.25 inch thick aluminum plates and the second configuration included one 0.5 inch thick aluminum plate. The finite element models for both configurations are shown in Figure 2.10.5-16 and Figure 2.10.5-17.

The stainless steel plates are connected by pipe elements representing the stainless steel bar and fusion welds. The aluminum plates are connected to the stainless steel bar via gap elements. A gap of 0.0 inches is used. The actual nominal gap is 0.06 inches.

Elastic material properties described in Section 2.10.5.2.2.B are used, and a uniform temperature of 450°F is applied.

The maximum stress intensities in the aluminum plates for configuration 1 and 2 are 0.84 ksi and 0.60 ksi respectively. The maximum stress intensities in the steel plates for configuration 1 and 2 are 1.45 ksi and 1.03 ksi respectively. The nodal stress intensity distributions in the stainless steel and aluminum plates are shown in Figure 2.10.5-18 through Figure 2.10.5-21 for both configurations.

Page 2.10.5-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.5.2.5 Basket Stress Analysis Conclusions Stresses in the stainless steel plates, aluminum plates, aluminum periphery plates, and aluminum rails are calculated for the NCT and HAC cask drop cases. The results for the side drop analysis are summarized in Table 2.10.5-4 through Table 2.10.5-6 and Table 2.10.5-7 through Table 2.10.5-9 for the NCT and HAC cases respectively. The results for the end drop analysis are summarized in the tables in Section 2.10.5.2.3.

The stresses in the stainless steel and aluminum plates are given in Section 2.10.5.2.4.

All stresses meet the stress criteria discussed in Section 2.10.5.2.2.B for both NCT and HAC evaluations.

2.10.5.3 TN-40 Fuel Basket Buckling Analysis 2.10.5.3.1 Analysis Approach Two techniques are available in the ANSYS program [1] for predicting the buckling load and buckling mode shape of a structure: nonlinear buckling analysis and eigenvalue (or linear) buckling analysis. Nonlinear buckling analysis is a more accurate approach since it can include features such as initial imperfections, material plastic behavior, gaps and large deflection response. This technique employs a nonlinear static analysis with gradually increasing loads to seek the load level at which structure becomes unstable.

Eigenvalue buckling analysis predicts the theoretical buckling strength of an ideal linear elastic structure. However, imperfections and nonlinearities prevent most real-world structures from achieving their theoretical elastic buckling strength. Thus, eigenvalue buckling analysis often yields unconservative results and is not recommended in actual engineering analysis [1]. Furthermore, this analysis is linear and can not account for the material plastic behavior.

An ANSYS nonlinear finite element analysis of the basket is conducted using large displacement and stress stiffening options with bilinear material properties to evaluate the plastic buckling loads. The five critical azimuth drop orientations analyzed are:

0° (load applied in the direction parallel to the basket vertical plates)

30° (load applied at 30° relative to the basket vertical plate direction)

45° (load applied at 45° relative to the basket vertical plate direction)

60° (load applied at 60° relative to the basket vertical plate direction)

90° (load applied in the direction perpendicular to basket vertical plates)

In order to calculate the buckling load, a three-dimensional finite element model of an 8 inch thick section of the basket is created. Figure 2.10.5-2 shows the model and the drop orientations.

Page 2.10.5-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The nodes between the steel baskets including the intermediate aluminum plates are coupled together in the out-of-plane direction so that they will bend in unison under surface pressure or other lateral loading to simulate through the thickness support provided by Boral plates. The aluminum plates are coupled together at their intersection. The fusion welds, connecting the fuel compartments and plates, are modeled by coupling nodes in all directions. The node couplings are shown in Figure 2.10.5-22. A sensitivity study on the effect of the nodal couplings and initial imperfections in the fuel boxes was performed and is provided in Section 2.10.5.5.

At the two cut faces, symmetry boundary conditions are applied (UY = ROTX = ROTZ =

0). The displacement constraints for the 0°, 30°, 45°, 60° and 90° side drop angles are shown in Figure 2.10.5-3 through Figure 2.10.5-7. For clarity, symmetry displacement constraints are not shown.

2.10.5.3.2 Buckling Analysis Loading Conditions The basket structure was analyzed for 0°, 30°, 45°, 60° and 90° side drops. Due to basket structure symmetry, these side drop azimuth orientations envelop all possible buckling modes.

Temperatures at the cross section where the maximum temperature occurred for the basket are taken from the NCT maximum environment of 100°F (Chapter 3). Figure 2.10.5-8 shows the temperature contour used in all analyses.

The load resulting from the fuel assembly weight is applied as pressure on the fuel compartment plates of the basket. For the 0° orientation, the pressure acts only on the horizontal plates, and for the 90° orientation, the pressure acted only on the vertical plates. But for the 30°, 45°, and 60° orientations, the pressure is divided into components that act on both horizontal and vertical plates of the basket. The pressures for the various orientations are calculated below for a 200g acceleration.

0° and 90° Drop Orientations Pressure for 1g, p = Fuel assembly weight / (Panel span x Panel length)

= 1300 lb. / (8.14 in. x 144 in.) = 1.109 psi Pressure for 200g = 200 x 1.109 = 222psi 30° Orientation Horizontal pressure for 1g = p sin30°

= 1.109 x 0.5 = 0.5545 psi Horizontal pressure for 200g = 222 x 0.5 = 111 psi Vertical pressure for 1g = p cos30°

= 1.109 x 0.866 = 0.9604 psi Vertical pressure for 200g = 222 x 0.866 = 192 psi Page 2.10.5-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 45° Orientation Pressure for 1 g = p cos45°

= 1.109 x 0.7071 = 0.7842 psi Pressure for 200 g = 222 x 0.7071 = 157 psi 60° Orientation Horizontal pressure for 1 g = p sin60°

= 1.109 x 0.866 = 0.9604 psi Horizontal pressure for 200 g = 222 x 0.866 = 192 psi Vertical pressure for 1 g = p cos60°

= 1.109 x 0.5 = 0.5545 psi Vertical pressure for 200 g = 222 x 0.5 = 111 psi The accelerations applied in each run are as follows.

Orientation Inertial Load (g) ax (g) ay (g) az (g) 0° 200 200 0 0 30° 200 173.21 0 -100.00 45° 200 141.42 0 -141.42 60° 200 100.00 0 -173.21 90° 200 0 0 -200 The pressure load distributions for the 0°, 30°, 45°, 60°, and 90° analyses are shown in Figure 2.10.5-23 through Figure 2.10.5-27.

2.10.5.3.3 Buckling Analysis and Results A maximum load of 200 g was applied to each analysis. The automatic time stepping option AUTOTS was activated. This option lets the program decide the actual size of the load sub-step for a converged solution. The last load step with a converged solution is the buckling load of the model.

The ANSYS input, buckling loads, and factors of safety for 0°, 30°, 45°, 60°, and 90° side drops are summarized in Table 2.10.5-11. Displacement patterns, at the last converged sub-step (buckling load) for the five cases are shown in Figure 2.10.5-28 through Figure 2.10.5-32. It may be seen that the displacements are not excessive at the last converged load step.

Page 2.10.5-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.5.4 Fusion Welds The testing program for the fusion welds ensures that the fusion weld is stronger than the base metal. Sections 2.10.5.2 and 2.10.5.3 calculates stresses in the base metal for normal condition of transport and hypothetical accident conditions and shows that the calculated stresses are below the stress limits. If the stresses in the base metal are below the stress limits and fusion welds are stronger than the base metal, basket integrity is maintained and the welds are qualified.

The testing program for the fusion welds is provided below:

The fusion spot welds that attach the stainless steel tubes or adjacent structural shapes shall be done by the GTAW fusion welding process, and be based on ANSI/AWS D1.3-89. This welding process shall produce a nugget of weld metal with a minimum 0.5 in. diameter weld shear area at the interface of the tubes and disk.

The GTAW machine welding parameters shall be preset and automated. For the production phase GTAW fusion spot welds, a one-hundred (100) percent visual inspection will verify the normality of the weld zone. In addition, a mechanical test of one test coupon from each welding machine used will verify proper machine settings and operation prior to the start of each working shift. The acceptance criteria will be failure of the base metal prior to failure of the weld area and a visual verification of a 1/2-inch diameter fused weld zone. Weld repairs will be made by following weld repair procedures (WRP).

The visual acceptance criteria for the fabricated welds shall be as follows:

Welds located up to 24" from the openings of the basket assemblies and directly visible shall be examined by direct visual inspection using the same acceptance criteria as the workmanship samples.

All other welds shall be examined by a remote visual inspection using mirrors and auxiliary lighting. This inspection shall verify the location, configuration and uniformity of the welds.

Excessive defects will be ground out by mechanical means and re-welded. Lack of penetration will be repaired by rewelding over the original weld zone. Reinspection by visual means to original standards is required. The automated GTAW fusion joining process shall be qualified using the guidelines of ASME B&PV Code Section IX and Section VIII Appendix 17.

As part of the weld qualification procedure, nine (9) test specimens shall be prepared and tested. All results shall be documented in a test report. All nine fusion spot welded specimens shall be prepared and shall be visually inspected for surface soundness, fusion and external nugget size. Three (3) of the nine specimens shall be sectioned and microetched.

Page 2.10.5-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The acceptance criteria are as follows. The fusion zone shall be sound with complete fusion along the bond line and a 3/32-inch weld penetration into the disk component. There shall be complete freedom from cracks along the bond line and the adjacent heat affected base metal. Small radial cracks at the center of the weld shall be considered non-relevant unless they exceed 1/8" in length as measured from the center of the weld to the end of the crack. The diameter of the weld nugget shall be at least 1/2 inch.

Undercut shall be considered non-relevant provided thorough fusion exists between the weld and base metal around the circumference of the weld and the length of the undercut does not exceed 3/16 inch.

Weld reinforcement shall range from 0.10 inch cavity (dish) to 1/16" to preclude interference with the test gage.

One of the three sectioned specimens shall undergo weld zone analysis for delta ferrite per NF-2433.1. Delta ferrite testing per NF-2433.1 shall be performed on weld samples prepared from each combination of sheet and disk heats to be spot welded together. The acceptance criterion is per NF-2433.2.

Peel tests shall be performed on three (3) other specimens and the acceptance criteria are as follows. The parent metal adjacent to the weld area must fail before the weld. The weld nugget at the bond line shall be free of defects and shall be at least 1/2- inch in diameter. Three (3) test specimens shall mechanically tested to failure. The base metal must fail before the weld zone.

2.10.5.5 Sensitivity Analysis 2.10.5.5.1 Couples Sensitivity Analysis The impact of node couplings which prevents the steel wall panels from buckling way from the aluminum plates was studied by modifying the finite element model described in Section 2.10.5.3. The node couplings between the basket panels were replaced with CONTACT52 [1] elements, this will allow the steel wall panels to buckle away from the aluminum plates while giving support if the buckle towards the aluminum plates. Weak spring elements were also created in parallel to the contact elements to prevent rigid body motion and help with convergence problems. Also, PIPE20 [1] elements were used to model the fusion welds. Figure 2.10.5-33 shows a representative panel with contacts and couples.

The boundary and loading conditions are described in Section 2.10.5.3 and were unchanged. Buckling analyses (as described in Section 2.10.5.3) were performed for the 0°, 30°, 45°, and 90° drop orientation.

Page 2.10.5-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The buckling loads are summarized in the table below. Figure 2.10.5-34 shows the displacement at the last converged load step (buckling load) for the bounding case, 0° side drop.

Buckling Load for Buckling Load for Contact Element Basket Side Drop Node Coupling Model(1) Model Orientation (g) (g) 0° 145.44 95.2 30° 88.54 96.6 45° 92.54 97.6 90° 115.15 96.6 Notes:

(1) Results from Section 2.10.5.3 (Table 2.10.5-11)

Changing node couplings to contact impacts the buckling loads of the 0° and 90° side drops, however the buckling loads of the bounding cases, 30° and 45°, are higher. In conclusion, the analysis with couples remains bounding.

2.10.5.5.2 Fuel Boxes Initial Imperfection Sensitivity Analysis The impact of initial imperfection in the fuel boxes was studied by applying an initial imperfection in the finite element model described in Section 2.10.5.5.1. Figure 2.10.5-35 shows the initial imperfection applied. Since 0° and 90° side drops are going to be affected by the initial imperfection, only the 0° orientation is analyzed in this study.

All conclusions from the 0° side drop case would be applicable to the 90° side drop also.

The boundary and loading conditions are described in Section 2.10.5.5.1 and were unchanged. Buckling analyses (as described in Section 2.10.5.3) was performed for the 0° drop orientation.

The buckling load for 0° side drops increases to 96.1g from 95.2g, which is marginal.

Displacement pattern, at the last converged sub-step (buckling load) is shown in Figure 2.10.5-36.

2.10.5.5.3 Tests Performed to Support Design of the TN-40 Basket Description of the Test In Revision 1 of the Prairie Island ISFSI (Appendix 4C of SAR, Docket 72-10) [4], tests were described that supported the design of the TN40 basket. Compression tests [7]

were performed to determine the basket panel strength when loaded in the 8.05 in.

direction (to simulate loading of a bottom outer radial panel of a basket during side impact). Each test panel was 8.05 in. long and 24 in. wide (the basket axial direction).

Panels were tested with weld spacings of 6 in., 8 in., and 12 in. in the axial direction.

The actual weld spacing is 8 in. Panels were tested at room temperature and at elevated temperature. The formula [8] for critical buckling stress in a rectangular plate under equal uniform compression is:

Page 2.10.5-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2

E t

= K 1 2 b where K = Factor depending on end supports and length to width ratio of the plate E = Modulus of Elasticity

= Poisson's Ratio t = thickness of the plate b = width of the plate K factors are typically higher for fixed supports than for pinned. The end supports for the basket panel are somewhere between a pinned-pinned support and a fixed-fixed support because there is edge rotational restraint provided by the corner of the stainless steel box sections and because of the aluminum plate continuation into adjacent panels.

Therefore, the tests were performed using a pinned-pinned configuration to yield conservatively low panel failure loads.

Test Results The test results from reference [7] are summarized below.

Weld Spacing Maximum Load Maximum Unit Load Test No. (in.) Temperature (°F) (lb, Total) (lb/in.)

A1 5 x 12 Room temp. 340,000 [7, Fig. 13] 14,166 B1 5x8 Room temp. 332,000 [7, Fig. 14] 13,833 C1 5x6 Room temp. 320,000 [7, Fig. 15] 13,333 A2 5 x 12 529 °F 253,000 [7, Fig. 20] 10,542 B2 5x8 405 °F 267,250 [7, Fig. 21] 11,135 C2 5x6 365 °F 261,500 [7, Fig. 22] 10,896 Correction Factor of the Test Compressive Load In Appendix 4C [4] and [7], the plate thickness used for the test specimens is 0.110 in.

for stainless steel plate and 0.256 in. for the aluminum plate. A correction factor is applied since the nominal plate thicknesses are 0.1 in. and 0.25 in. for fuel compartment and aluminum plates, respectively.

Page 2.10.5-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Correction Factor, F Lb13 3

+ Lb2 + Ah12 + Ah22 12 12

= 3 Lb3 Lb 3

+ 4 + Ah32 + Ah42 12 12 0.2563 0.113

+ + (0.256)(0.256 ) 2 + (0.11)(0.256 + 0.11 ) 2 12 12 2 2

=

0.25 3 0.13

+ + (0.25)(0.25 ) 2 + (0.1)(0.25 + 0.1 ) 2 12 12 2 2 0.016343

= = 1.1435 0.014292 where b1, b2, h1, and h2 are plate thickness and neutral axis offset for the test specimen aluminum and stainless steel, respectively. Similarly b3, b4, h3, and h4 are the plate thickness and neutral axis offset for the TN40 basket panel and A = length (1 in.) x thickness.

The minimum test buckling load is 10,542 lb/in. (at 529 °F). Applying the correction factor, the calculated test buckling load based on the nominal plate thicknesses gives:

Test buckling load = 10,542 / 1.1435 = 9,219 lb/in.

Compressive Load Calculated by the ANSYS Model To calculate the compressive load during a 75 g accident side drop event, a finite element model was created in ANSYS [1]. The test setup of Appendix 4C has the basket panel supported by a pin-pin connection. Therefore, the TN40 basket panels are modeled with pin-pin supports. A 90° drop orientation, which results in the maximum compressive load in the bottom-most fuel compartment, is analyzed in the calculation.

The purpose of the analysis is to determine the maximum compressive load in the TN40 basket panels. To simplify the analysis the sandwich (aluminum + steel) panels of the basket are modeled as a single plate using ANSYS Shell43 elements 0.5 in. thick.

Furthermore, the panels are modeled in such a way that there is no moment transfer through any of the fuel compartments. The basket is modeled as an 8 in. section in the axial direction. Figure 2.10.5-37 shows the finite element model of the representative TN40 basket.

The basket is supported at the rail locations appropriate to the drop orientation.

Symmetric boundary conditions are applied to the cut boundary locations.

Elastic material properties are used. The density of the material is adjusted such that the weight of the single plate model is identical to the basket weight.

Page 2.10.5-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The load resulting from the fuel assembly weight was applied as pressure on the fuel compartment plates of the basket.

A linear elastic side drop analysis of the basket is performed using ANSYS. The maximum compressive force in the bottom most fuel compartment wall is calculated to be 42,530 lb. Since the basket is modeled as an 8 in. section, the unit load in the fuel compartment wall is 42,530/8 = 5,316 lb/in.

Conclusion The maximum compressive load for the 75 g accident side drop load is 5,316 lb/in. The buckling load from the test specimen is 9,219 lb/in. Therefore, the factor of safety against buckling of the panel is 1.73 (9,219/5,316). Based on the test, it can be concluded that the basket can withstand up to 130 g (75 x 1.73) before reaching the buckling load.

2.10.5.6 Summary Nonlinear analyses with bilinear material properties and small deflections were performed in ANSYS for the critical azimuth side drop orientations to determine the membrane and membrane plus bending stresses in all basket components. It was shown that the stresses at 75 g for a 30 foot drop are below allowable stress limits. For these analyses, steel tubes, including the intermediate aluminum plates, are connected together in the out-of-plane direction so that they will bend in unison under surface pressure or other lateral loading to simulate through-the-thickness support provided by the Boral plates.

The same model was used to determine the critical buckling load for the basket, except large displacement and stress stiffening options were used. The buckling analyses are reported in Section 2.10.5.3. From these analyses a minimum buckling load of 88.54 g was determined. In addition, several sensitivity analyses were performed (Section 2.10.5.5) to investigate the effect of modeling assumptions and geometrical imperfections.

The first sensitivity analysis investigated the effect of connecting the steel tubes and the intermediate aluminum plates in the out-of-plane direction. The steel tubes were allowed to separate from each other to correctly simulate the buckling behavior. The results showed that even though the buckling loads when the basket is oriented in 0 and 90 degrees direction are lower than previously calculated, the minimum buckling load of 88.54 g is still bounding.

The second sensitivity analysis investigated the effect of initial imperfections. An initial imperfection was applied to all steel tubes and the buckling load was calculated for the bounding basket orientation. The results showed that the initial imperfection does not have an effect on the calculated buckling load.

Page 2.10.5-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Furthermore, compression tests were performed to determine the basket panel strength during side impact (Section 2.10.5.5.3) at room temperature as well as elevated temperature. The results showed a safety factor of 1.73 with respect to the design load of 75 g.

ANSYS buckling analyses performed in Section 2.10.5.3 and Section 2.10.5.5 for an 8.0 inch sector assumes temperatures at the hottest section for the 100 °F ambient conditions. The minimum calculated buckling load of 88.5 g provides sufficient safety factors for all loading conditions (basket baseline g loads are provided in Section 2.7.1 of Chapter 2) except for the slapdown impact when the ambient condition is -20 °F. For the -20 °F ambient conditions, the top and bottom portions of the basket are at lower temperatures than the temperatures used in the buckling analyses and lower temperature will increase the buckling load. The average temperatures in the basket periphery for each condition are provided in the table below. The temperature dependent material properties for SA-240 Gr. 304 and SB-209 6061-T651 at these temperatures are interpolated from data provided in Table 2.10.5-1. It is seen that the Youngs modulus for SA-240 Gr. 304 and SB-209 6061-T651 increase by 2.2% and 4.4%, respectively, when the temperature decreases from 330 °F to 210 °F. Also the yield strength for SA-240 Gr. 304 and SB-209 6061-T651 increase by 15.7% and 31.0%, respectively, when the temperature decreases from 330 °F to 210 °F.

The effect of basket temperature on the buckling load is evaluated using the results from limit load tests presented in Section 2.10.5.5.3. The limit load tests were performed at room temperature (70 °F) as well as elevated temperatures (365 to 529 °F). It is seen that because of the higher Youngs modulus and yield strength at lower temperatures, the load at collapse for the tests performed at room temperature is much higher then the load at collapse for higher temperature. Using the test results from Section 2.10.5.5.3:

Average load at collapse at room temperature: 13,777 lb/in.

Average load at collapse at elevated temperature: 10,858 lb/in.

Average elevated temperature: 433 °F Room temperature: 70 °F The buckling load is 27% higher for the room temperature tests than at elevated temperature. Assuming a linear relationship the buckling load would increase by 9.4%

for a 126 °F decrease in temperature. Therefore, the adjusted buckling load at 210 °F for the -20 °F ambient condition is 96.9 g (88.54 x 1.094).

Page 2.10.5-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Therefore, the safety factors for buckling load with its respective g load are:

Average Average Temperature in Lowest Baselin Temperature in the the Basket Periphery Buckling eG Ambient Drop Basket Periphery(1) used in the Analysis G Load Load Safety Condition Orientation (°F) (°F) (g) (g) Factor Side drop 336 336 88.5 55 1.61 100 °F Slapdown 312 336 88.5 63 1.40 Side drop 234 336 88.5 63 1.40

-20 °F Slapdown 210 210 96.9 72 1.35 (1) The average temperatures in the basket periphery are calculated from the ANSYS results files generated in the Section 3.4 NCT thermal analysis.

Based on the above basket analyses, it is shown that the calculated basket stresses meet the ASME Code allowables. In addition, the minimum safety factor for buckling is 1.35. This buckling safety factor is considered sufficient to assure the structural performance of the basket. The following discussion is provided in support of this conclusion:

1) The factors of safety required by the ASME code for stainless steel (1.41 to 2.21 as calculated following NUREG/CR-6322 [9]) are based on elastic analysis. As discussed in NUREG/CR-6322, the magnitude of these factors of safety are intended to provide additional conservatism due to the following factors:

a. The analysis approach taken in NUREG/CR-6322 is based on the design practice where the entire structure is designed by sizing the individual members of the assemblies. The compressive load in a member will influence the critical buckling load of not only the member itself but also other adjacent members that are connected to the same structural joint. If the basket design is based on one individual member, then the individual member interactions with other members will not be included.

A full 360 degree sector of the basket model with elastic-plastic material and large deflection effects is used to calculate the buckling limit. The full 360 degree basket model takes into account all the interactions among all the basket members.

b. A real member may have imperfections that include initial curvature of a member, eccentric loads on initially straight member, and residual stresses due to forming or assembly. These imperfections tend to make the actual failure load lower than the theoretical critical load.

Page 2.10.5-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The configuration and analysis methodology used for the TN-40 basket tend to mitigate these concerns. First, a sensitivity study (SAR Section 2.10.5.5.2) was performed to evaluate the impact of geometrical imperfections. Based on the study, the buckling load remained the same. The study concluded that for this type of basket design (composite structural with fusion welds), the buckling effect due to initial imperfections is minimal. Second, the fuel compartment panels are part of a complete tube such that pressure loads on a horizontal panel will produce bending loads and deflections in the adjoining vertical panels. This in effect imposes an eccentric load that is addressed in the determination of the buckling load.

2) Buckling evaluations are also addressed in ASME Section III, Division 3, Subsection WD [10]. The rules in Subsection WD originate from Subsection NG (Core Support Structures) and Subsection NF (Supports) and are intended to be used for transportation and storage basket design. The buckling analysis methodology is described in Subsection WD-3229. Subsections WD-3229.2 and 3229.3 describe the analysis of rectangular plates under compressive loading. The allowable compressive stresses given are as follows:

a. Normal condition: Fnormal = 0.5 Fcritical

b. Accident condition: Faccident = 1.5 x Fnormal = 1.5 x 0.5 Fcritical = 0.75 x Fcritical The safety factor for the accident condition is therefore 1/0.75 = 1.33

3) The yield strengths of stainless steel and aluminum increase at high strain rates comparable to those resulting from a 30 foot drop. The resulting yield stress for stainless material (304/304L) at 300° F is expected to increase approximately 16% to 31% [11]. Thus the basket buckling load and resulting safety factor will increase if the strain rate effects are included in the analyses.

4) A conservative fuel weight is also used in the analysis. The total length of the fuel assembly is 161.3 in. as shown in Chapter 1, Section 1.2.3 (page 1-8). For the basket buckling analysis, the fuel weight is distributed over 144 in. of the basket (based on the active fuel length, Section 2.10.5.3.2). During the slapdown load case (bounding safety factor), the maximum g load occurs at either the top or bottom end of the basket, depending on drop orientation. However, the weights of the fuel assembly at top region (fuel-gas plenum zone and top end fitting zone, 17.68 kg (38.98 lbs), Table 5-4 of Chapter 5) and bottom region (bottom end fitting zone, 7.89 kg (17.39 lbs), Table 5-4 of Chapter 5) is much lower than the fuel assembly weight of an equal length in the active fuel region. Using the same approach as used in SAR Section 2.10.5.3.2, the fuel pressure load on the basket panels at the ends of the basket are:

Basket length = (total length - active fuel region) / 2

= (160 - 144) / 2 = 8 in Pressure = (bounding weight of the top or bottom region) / compartment area

= 38.98 / (8.14 x 8) = 0.6 psi Page 2.10.5-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 This pressure (0.6 psi) is much smaller than the pressure in the active fuel region (1.109 psi). With this smaller pressure (0.6 psi), the safety factor for the buckling load due to slapdown is approximately 2.5 (1.35 x 1.109/0.6).

In view of the discussion above the calculated minimum safety factor of 1.35 is sufficient to ensure that the basket is capable of withstanding the accident impact loadings.

Page 2.10.5-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.5.7 References

1. ANSYS Engineering Analysis System User's Manual, Releases 8.0 and 10.0.

2. ASME Boiler and Pressure Vessel Code, 1989,Section III, Subsection NB, NF &

Appendices;Section VIII, Divs I &2.

3. Aluminum Standards and Data, The Aluminum Association, Inc., 1976.

4. Prairie Island ISFSI Technical Specification and Safety Analysis Report, Revision 1, 1991.

5. An Assessment of Stress-Strain Data Suitable for Finite Element Elastic-Plastic Analysis of Shipping Containers NUREG/CR-0481, SAND77-1872.

6. Kaufman, J. Gilbert, Properties of Aluminum Alloys: Tensile, Creep, and Fatigue Data at High and Low Temperatures, 1999.

7. Scavuzzo, R. J., Lam, P. C., Gau, J. S., Buckling Tests of Fusion Welded Composite Stainless Steel Aluminum Plates, Dept. of Mechanical Engineering, The University of Akron, May, 1990.

8. Young, Warren C. and Budynas, Richard G., Roark's Formulas for Stress and Strain, Seventh Edition, McGraw-Hill, New York, 2002.

9. Buckling Analysis of Spent Fuel Basket, NUREG/CR-6322, May 1995.

10. ASME Boiler and Pressure Vessel Code,Section III, Division 3, Subsection WD Internal Support Structures, Draft, November 2010.

11. Dana K. Morton, Robert K. Blandford, Spencer D. Snow, Impact Testing of Stainless Steel Material at Cold Temperatures, PVP2008-61215.

Page 2.10.5-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-1 Material Properties for TN-40 Fuel Basket Temperature Sy Su E m Density Part Material (F) (psi) (psi) (psix106) (in/in/F) (lb/in3) 70 35,000 42,000 10.0 - 0.1085 SB-209, 300 27,400 31,700 9.2 13.22x10-6 0.1085 Al Plates 6061-400 13,300 17,700 8.7 13.52x10-6 0.1085 T651 500 4,375 7,000 8.1 13.82x10-6 0.1085 70 30,000 75,000 28.3 - 0.29 Steel SA-240, 300 22,500 66,000 27.0 9.0x10-6 0.29 Boxes Gr. 304 400 20,700 64,400 26.5 9.19x10-6 0.29 500 19,400 63,500 25.8 9.37x10-6 0.29 70 35,000 42,000 10.0 - 0.098 SB-209, Outer 300 27,400 31,700 9.2 13.22x10-6 0.098 6061-Plates 400 13,300 17,700 8.7 13.52x10-6 0.098 T651 500 4,375 7,000 8.1 13.82x10-6 0.098 70 35,000 42,000 10.0 - 0.098 Al SB-209, 300 27,400 31,700 9.2 13.22x10-6 0.098 Peripher 6061-y Plates 400 13,300 17,700 8.7 13.52x10-6 0.098 T651 500 4,375 7,000 8.1 13.82x10-6 0.098 Notes:

Material Properties are obtained from ASME code Section III Appendices [2]. Aluminum material properties at elevated temperatures are taken from aluminum standards and data [3].

5% of the elastic modulus used as the tangent modulus

Since the Boral plates were not included in the analysis, the weight of the plates is included in the weight of the aluminum plates. The density of the aluminum plates is adjusted to match the total weight of the Boral and aluminum plates.

Page 2.10.5-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-2 Basket Structural Allowable Stresses, NCT Temperature Pm (Sm or S) Pm + Pb (1.5Sm or 1.5S)

Material (F) (psi) (psi) 70 14,000 21,000 SB-209, 300 11,300 16,950 6061-T651 400 4,400 6,600 500 1,750 2,625 70 20,000 30,000 SA-240, 300 20,000 30,000 Gr. 304 400 18,700 28,050 500 17,500 26,250 Notes:

Sm obtained from ASME code Section VIII, Division 2; S obtained from Section III or the lower of Su/4 or 2/3 Sy Table 2.10.5-3 Basket Structural Allowable Stresses, HAC Temperature Pm (0.7 Su) Pm + Pb (0.9 Su)

Material (F) (psi) (psi) 70 29,400 37,800 SB-209, 6061- 300 22,190 28,530 T651 400 12,390 15,930 500 4,900 6,300 70 52,500 67,500 300 46,200 59,400 SA-240, Gr. 304 400 45,080 57,960 500 44,450 57,150 Page 2.10.5-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-4 NCT 0° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (20g) Stress Location Category (F)

(ksi) (ksi)

Pm 5.8 310 19.9 Max Stress S.S. Boxes Pm + Pb 6.7 310 29.8 and Rails Pm 5.8 333 19.6 Min F.S.

Pm + Pb 6.7 310 29.8 Pm 4.5 316 10.2 Max Stress Aluminum Pm + Pb 4.7 316 15.3 Plates Pm 4.5 316 10.2 Min F.S.

Pm + Pb 2.9 437 5.1 Pm 1.3 298 11.3 Aluminum Max Stress Pm + Pb 1.9 339 12.9 Periphery Plates Pm 1.1 326 9.5 Min F.S.

Pm + Pb 1.9 339 12.9 Pm 0.8 288 11.4 Max Stress Aluminum Pm + Pb 1.7 288 17.2 Outer Plates Pm 0.8 288 11.4 Min F.S.

Pm + Pb 1.7 288 17.2

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Page 2.10.5-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-5 NCT 45° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (20g) Stress Location Category (F)

(ksi) (ksi)

Pm 4.6 351 19.3 Max Stress S.S. Boxes Pm + Pb 6.9 317 29.7 and Rails Pm 4.6 351 19.3 Min F.S.

Pm + Pb 6.9 317 29.7 Pm 3.3 308 10.8 Max Stress Aluminum Pm + Pb 4.5 251 17.8 Plates Pm 3.2 316 10.2 Min F.S.

Pm + Pb 3.7 440 5.0 Pm 0.9 298 11.3 Aluminum Max Stress Pm + Pb 2.5 323 14.6 Periphery Plates Pm 0.8 326 9.5 Min F.S.

Pm + Pb 2.5 323 14.6 Pm 0.9 251 11.8 Max Stress Aluminum Pm + Pb 1.4 251 17.8 Outer Plates Pm 0.9 251 11.8 Min F.S.

Pm + Pb 1.4 251 17.8

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Page 2.10.5-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-6 NCT 90° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (20g) Stress Location Category (F)

(ksi) (ksi)

Pm 5.5 365 19.2 Max Stress S.S. Boxes Pm + Pb 8.7 308 29.8 and Rails Pm 5.5 365 19.2 Min F.S.

Pm + Pb 8.7 308 29.8 Pm 3.9 316 10.2 Max Stress Aluminum Pm + Pb 4.6 367 10.0 Plates Pm 3.8 320 9.9 Min F.S.

Pm + Pb 4.2 440 5.0 Pm 0.5 288 11.4 Aluminum Max Stress Pm + Pb 2.2 328 14.1 Periphery Plates Pm 0.4 337 8.8 Min F.S.

Pm + Pb 1.9 360 10.7 Pm 0.7 251 11.8 Max Stress Aluminum Pm + Pb 0.9 251 17.8 Outer Plates Pm 0.7 251 11.8 Min F.S.

Pm + Pb 0.9 251 17.8

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Page 2.10.5-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-7 HAC 0° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (75g) Stress Location Category (F)

(ksi) (ksi)

Pm 24.7 310 46.1 Max Stress S.S. Boxes Pm + Pb 29.2 310 59.3 and Rails Pm 24.7 310 46.1 Min F.S.

Pm + Pb 29.2 310 59.3 Pm 16.7 316 20.6 Max Stress Aluminum Pm + Pb 17.3 245 30.7 Plates Pm 16.7 316 20.6 Min F.S.

Pm + Pb 8.6 437 12.3 Pm 4.7 298 22.3 Aluminum Max Stress Pm + Pb 7.3 339 23.7 Periphery Plates Pm 4.3 326 19.6 Min F.S.

Pm + Pb 7.3 339 23.7 Pm 3.1 288 22.5 Max Stress Aluminum Pm + Pb 6.4 288 29.0 Outer Plates Pm 3.1 288 22.5 Min F.S.

Pm + Pb 6.4 288 29.0

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Page 2.10.5-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-8 HAC 45° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (75g) Stress Location Category (F)

(ksi) (ksi)

Pm 16.7 351 45.6 Max Stress S.S. Boxes Pm + Pb 29.0 308 59.3 and Rails Pm 16.7 351 45.6 Min F.S.

Pm + Pb 29.0 308 59.3 Pm 12.3 308 21.4 Max Stress Aluminum Pm + Pb 15.4 382 18.3 Plates Pm 12.2 313 20.9 Min F.S.

Pm + Pb 14.1 415 14.5 Pm 3.4 298 22.3 Aluminum Max Stress Pm + Pb 8.6 323 25.6 Periphery Plates Pm 3.1 326 19.6 Min F.S.

Pm + Pb 8.6 323 25.6 Pm 3.5 251 23.7 Max Stress Aluminum Pm + Pb 5.3 251 30.5 Outer Plates Pm 3.5 251 23.7 Min F.S.

Pm + Pb 5.3 251 30.5

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Page 2.10.5-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-9 HAC 90° Side Drop, Basket Stress Analysis Results Stress Allowable Stress* Stress Temperature Component (75g) Stress Location Category (F)

(ksi) (ksi)

Max Stress Pm 21.1 316 46.0 S.S. Boxes Pm + Pb 25.7 308 59.3 and Rails Min F.S. Pm 21.1 316 46.0 Pm + Pb 25.7 308 59.3 Max Stress Pm 14.4 316 20.6 Aluminum Pm + Pb 17.1 367 20.1 Plates Min F.S. Pm 14.1 320 20.2 Pm + Pb 15.4 403 15.6 Max Stress Pm 2.0 288 22.6 Aluminum Pm + Pb 8.4 328 25.1 Periphery Plates Min F.S. Pm 1.6 337 18.6 Pm + Pb 7.3 358 21.2 Max Stress Pm 2.6 251 23.7 Aluminum Pm + Pb 3.5 251 30.5 Outer Plates Min F.S. Pm 2.6 251 23.7 Pm + Pb 3.5 251 30.5

Since the allowable stress is based on the temperature of the location where the stress is occurring, two locations are reported: Max Stress and Min F.S., where:

Max Stress: Location where maximum stress is occurring regardless of the allowable stress at that location Min F.S.: Location where the factor of safety is minimum based on the allowable stress at that location Table 2.10.5-10 DELETED Page 2.10.5-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.5-11 Fuel Basket Buckling Analysis Results Maximum Load used in Analyses Basket Side Maximum Vertical Horizontal Last Converged Drop Acceleration Pressure Pressure Load Orientation (g) (psi) (psi) (g) 0 200 222 0 145.44 30 200 192 111 88.54 45 200 157 157 92.54 60 200 111 192 92.54 90 200 0 222 115.15 Page 2.10.5-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-1 Typical TN-40 Basket Fuel Compartment Plate Page 2.10.5-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-2 TN-40 Basket Finite Element Model Including Drop Orientations Page 2.10.5-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-3 Basket Finite Element Model Displacement Constraints - 0° Side Drop Page 2.10.5-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-4 Basket Finite Element Model Displacement Constraints - 30° Side Drop Page 2.10.5-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-5 Basket Finite Element Model Displacement Constraints - 45° Side Drop Page 2.10.5-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-6 Basket Finite Element Model Displacement Constraints - 60° Side Drop Page 2.10.5-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-7 Basket Finite Element Model Displacement Constraints - 90° Side Drop Page 2.10.5-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-8 Basket Finite Element Model Temperature Boundary Condition Page 2.10.5-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-9 Basket Finite Element Model Applied Pressures - 0° Drop, NCT Page 2.10.5-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-10 Basket Finite Element Model Applied Pressures - 45° Drop, NCT Page 2.10.5-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-11 Basket Finite Element Model Applied Pressures - 90° Drop, NCT Page 2.10.5-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-12 NCT 45° Side Drop - S.S. Plates - Membrane plus Bending Stress Intensity Page 2.10.5-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-13 NCT 45° Side Drop - Al. Plates - Membrane plus Bending Stress Intensity Page 2.10.5-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-14 NCT 45° Side Drop - Al. Periphery Plates -

Membrane plus Bending Stress Intensity Page 2.10.5-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-15 NCT 45° Side Drop - Al. Outer Plates -

Membrane plus Bending Stress Intensity Page 2.10.5-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-16 Thermal Stress Analysis Finite Element Model - Configuration 1 Page 2.10.5-49

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-17 Thermal Stress Analysis Finite Element Model - Configuration 2 Page 2.10.5-50

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-18 Thermal Stress Analysis - Configuration 1 -

Maximum Aluminum Stress Intensity Page 2.10.5-51

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-19 Thermal Stress Analysis - Configuration 1 -

Maximum Stainless Steel Stress Intensity Page 2.10.5-52

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-20 Thermal Stress Analysis - Configuration 2 -

Maximum Aluminum Stress Intensity Page 2.10.5-53

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-21 Thermal Stress Analysis - Configuration 2 -

Maximum Stainless Steel Stress Intensity Page 2.10.5-54

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-22 Fuel Basket Buckling Analysis Finite Element Model Node Couplings Page 2.10.5-55

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-23 Basket Buckling Analysis Loading Boundary Conditions - 0° Side Drop Page 2.10.5-56

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-24 Basket Buckling Analysis Loading Boundary Conditions - 30° Side Drop Page 2.10.5-57

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-25 Basket Buckling Analysis Loading Boundary Conditions - 45° Side Drop Page 2.10.5-58

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-26 Basket Buckling Analysis Loading Boundary Conditions - 60° Side Drop Page 2.10.5-59

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-27 Basket Buckling Analysis Loading Boundary Conditions - 90° Side Drop Page 2.10.5-60

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-28 Fuel Basket Deformation at Buckling Load - 0° Side Drop Page 2.10.5-61

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-29 Fuel Basket Deformation at Buckling Load - 30° Side Drop Page 2.10.5-62

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-30 Fuel Basket Deformation at Buckling Load - 45° Side Drop Page 2.10.5-63

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-31 Fuel Basket Deformation at Buckling Load - 60° Side Drop Page 2.10.5-64

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-32 Fuel Basket Deformation at Buckling Load - 90° Side Drop Page 2.10.5-65

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-33 Fuel Compartment Interface Elements used in Sensitivity Study Page 2.10.5-66

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-34 Basket Deformation at Buckling Load for 0 Degree Side Drop Page 2.10.5-67

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-35 Initial Imperfections Imposed (Deformation is exaggerated for effect however the contour scale is correct)

Page 2.10.5-68

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-36 Basket Deformation at Buckling Load for 0 Degree Side Drop with Initial Imperfections Page 2.10.5-69

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.5-37 Basket Finite Element Model to Calculate the Maximum Compressive Load Page 2.10.5-70

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.6 DYNAMIC LOAD FACTOR FOR BASKET DROP ANALYSIS TABLE OF CONTENTS 2.10.6.1 Introduction........................................................................ 2.10.6-1 2.10.6.2 Modal Analysis of Basket Side Drop Loading Condition .... 2.10.6-1 2.10.6.3 Frequency of Basket due to Basket End Drop Loading Condition ........................................................................... 2.10.6-3 2.10.6.4 Dynamic Load Factor Calculations .................................... 2.10.6-3 2.10.6.5 References ........................................................................ 2.10.6-5 LIST OF FIGURES Figure 2.10.6-1 TN-40 Basket Finite Element Model ....................................... 2.10.6-6 Figure 2.10.6-2 Displacement Boundary Condition - 0° Side Drop .................. 2.10.6-7 Figure 2.10.6-3 0° Side Drop - 1st Mode ......................................................... 2.10.6-8 Figure 2.10.6-4 45° Side Drop - 1st Mode ....................................................... 2.10.6-9 Figure 2.10.6-5 90° Side Drop - 1st Mode ..................................................... 2.10.6-10 Figure 2.10.6-6 DLF Calculation Relationship................................................ 2.10.6-11 Figure 2.10.6-7 DELETED ............................................................................. 2.10.6-12 Figure 2.10.6-8 DELETED ............................................................................. 2.10.6-12 Page 2.10.6-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 DYNAMIC LOAD FACTOR FOR BASKET DROP ANALYSIS 2.10.6.1 Introduction This appendix presents the modal analysis of the TN-40 fuel basket. The TN-40 basket is analyzed for 30 foot end drop and side drop accidents in Appendix 2.10.5 using equivalent static methods. The equivalent static loads for the drop evaluations of the TN-40 basket are determined by multiplying the baseline rigid body accelerations (as discussed in Section 2.7.1) by the corresponding dynamic load factor (DLF). The DLF is a function of the rise time of the applied load, the duration of the load, the shape of the load, and the modal frequencies of the structure. The purpose of this analysis is to determine the fundamental frequencies of the basket which have the most significant effect on the response of the basket to the 30 foot side impact. Using the fundamental frequencies of the basket structure, the DLF is determined from the curve shown in Figure 2.10.6-6 which is taken from the NUREG/CR-3966 [1]. The results give the DLFs for a half-sine-wave as a function of the ratio of the impulse duration to the natural period of the structure. The half sine wave is used because it gives a reasonable approximation of the actual load experienced by the cask during a drop event.

2.10.6.2 Modal Analysis of Basket Side Drop Loading Condition Finite Element Model Modal analyses were run using the ANSYS [2] finite element model described in Appendix 2.10.5. The model was modified such that the fuel weight is included as mass instead of pressure. The basket finite element model is shown on Figure 2.10.6-1.

Material Properties The stainless steel boxes are constructed of SA-240, Gr. 304 stainless steel. The aluminum plates, outer plates and basket periphery plates are constructed of SB-209, 6061-T651 aluminum alloy. The calculated maximum temperature in the basket is 444F (Chapter 3, Table 3-1), material properties at 450F are used. The following material properties are used in the analysis [5][6].

For SA-240, Gr. 304 at 450F:

E = 26.15 x 106 psi NU = 0.3

= 0.29/386.4 lbm/in3 For SB-209, 6061-T651 at 450F:

E = 8.4 x 106 psi NU = 0.3

= 0.098/386.4 lbm/in3 Page 2.10.6-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The density of the aluminum plates was adjusted to match the total weight of the Boral and aluminum plates:

= AL x (WAL + WBoral) / (WAL)

= (0.098 / 386.4) x (5402.6 + 579.5) / (5402.6)

= 2.81 x 10-4 lbm/in3 The weight of the fuel (1300 lbs) was added to the bottom or right surface of the steel basket depending on the drop orientation.

For 0 and 90 Side Drop = st + [(1300 / 386.4) x (8 / 144) / (8 x 8.14 x 0.09)] =

0.033 lbm/in3 For 45 Side Drop = st + [0.7071 (1300 / 386.4) x (8 / 144) / (8 x 8.14 x 0.09)] =

0.023 lbm/in3 Boundary Conditions The bottom half of the basket perimeter is constrained in the direction parallel to the drop angle and the entire perimeter is constrained in the perpendicular direction. These boundary conditions were chosen to eliminate modes of vibration that are incompatible with the physics of the drop. For instance, side to side modes are not important because they are restrained by the basket wall and more importantly, because they will have no modal weight in the drop direction and therefore will not be activated by the drop. Typical boundary conditions for the 0° modal analysis are shown on Figure 2.10.6-2.

Results of the Modal Analysis The first five frequencies for each drop angle are summarized in the following table.

The deformed shapes of the first mode of each drop are shown in Figure 2.10.6-3 to Figure 2.10.6-5.

Results Summary - Natural Frequencies Frequency (Hz)

Mode 0 Drop 45 Drop 90 Drop 1 148.3 165.9 148.7 2 151.3 166.1 148.7 3 154.0 168.3 151.4 4 159.0 171.5 152.2 5 161.0 171.9 154.3 Page 2.10.6-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.6.3 Frequency of Basket due to Basket End Drop Loading Condition The fundamental natural frequency of a simply supported cylindrical shell under axial vibration simplifies to that of a uniform beam, free axially at both ends. The fundamental natural frequency of a uniform beam free at both ends, under longitudinal vibrations is as follows [4].

1 E 2 f1 = 1 2l Where l =

The average mass density, , is calculated the following way, the stiffnes of the aluminum and Boral plates are ignored:

Total weight for steel components = 6,608.8 lbs Total volume for steel components = 6,608.8 / 0.29 = 22,789.0 in.3 Total weight for aluminum/ Boral components = 8,083.7 lbs Average weight density = (6,608.8 + 8,083.7) / 22,789 = 0.645 lb in.-3 Average mass density, = 0.645 / 386.4 = 0.00167 lbm. in.-3 l = 160 in.

Therefore, 1

26.15 10 6 2 f1 = = 391 .0 Hz 2 ( 160 ) 0.00167 The stiffness of the aluminum and Boral plates are ignored which will increase the natural frequencies.

2.10.6.4 Dynamic Load Factor Calculations The DLF is computed for the end and side drops. The impact duration from the TN-40 1/3 scale model drop test (Appendix 2.10.9) is used.

The DLF calculation procedure in Reference [1] is based on the natural time period T (or 1/natural frequency) of the structure, and the duration and shape of the impact impulse.

Page 2.10.6-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The DLF of the TN-40 basket are based on a half sine wave shape impulse and cask impact impulse durations (t) of 0.0525 and 0.075 sec.

The DLF calculated for side and end drops are shown in the following table, using Figure 2.10.6-6 and the lowest natural frequency calculated in Sections 2.10.6.2 and 2.10.6.3.

Dynamic Load Factor Calculations Natural Natural Time Impulse DLF Drop Frequency Period, T Duration(1), t Ratio t/T (from Orientation (Hz) (Sec.) (Sec) Figure 2.10.6-6)

End Drop 391.0 0.0026 0.0525 20.2 < 1.08 Side Drop 148.3 0.0067 0.075 11.1 < 1.08 Notes:

(1) Impact durations from the TN40 1/3 scale impact limiter testing are 0.0175 and 0.025 sec for end and side drops, respectively. These are equivalent to 0.0525 and 0.075 sec for the full scale impact limiter. The time history of the TN40 1/3 Scale impact limiter End and Side Drop are shown in Appendix 2.10.9, Figures 2.10.9-24 and 2.10.9-10 respectively.

Page 2.10.6-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.6.5 References

1. NUREG/CR-3966, Methods for Impact Analysis of Shipping Containers, 11/1987.

2. ANSYS Engineering Analysis System Users Manual, Rev. 8.0.

3. DELETED

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

5. ASME Boiler and Pressure Vessel Code, 1989,Section III & Appendices;Section VIII, Divs 1 & 2.

6. Aluminum Standards and Data, The Aluminum Association, Inc., 1976.

Page 2.10.6-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-1 TN-40 Basket Finite Element Model Page 2.10.6-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-2 Displacement Boundary Condition - 0° Side Drop Page 2.10.6-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-3 0° Side Drop - 1st Mode Page 2.10.6-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-4 45° Side Drop - 1st Mode Page 2.10.6-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-5 90° Side Drop - 1st Mode Page 2.10.6-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 (Reproduced from Reference [1])

Figure 2.10.6-6 DLF Calculation Relationship Page 2.10.6-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.6-7 DELETED Figure 2.10.6-8 DELETED Page 2.10.6-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 2.10.7-i and 2.10.7-ii and Pages 2.10.7-1 through 2.10.7-36 Withheld Pursuant to 10 CFR 2.390 Page 2.10.7-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.8 STRUCTURAL EVALUATION OF THE IMPACT LIMITERS TABLE OF CONTENTS 2.10.8.1 Introduction........................................................................ 2.10.8-1 2.10.8.2 Design Description ............................................................ 2.10.8-1 2.10.8.3 Design Criteria ................................................................... 2.10.8-3 2.10.8.4 Analysis of the HAC 30 Foot Free Drop ............................ 2.10.8-4 2.10.8.5 Analysis for One Foot Drop Normal Condition of Transport ........................................................................... 2.10.8-7 2.10.8.6 Impact Limiter Attachment Analysis .................................. 2.10.8-8 2.10.8.7 Summary of ADOC Results Used for Structural Analysis ........................................................................... 2.10.8-20 2.10.8.8 Summary Description of ADOC Computer Code ............. 2.10.8-21 2.10.8.9 References ...................................................................... 2.10.8-32 Page 2.10.8-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 2.10.8-1 Mechanical Properties of Wood and Wood Adhesive ........... 2.10.8-33 Table 2.10.8-2 Typical Wood Material Properties ......................................... 2.10.8-34 Table 2.10.8-3 First Impact Maximum Inertia g-Load versus Initial Angle of Impact for 30 Foot Drop, Using Maximum Wood Crush Stress Properties .................................................................. 2.10.8-35 Table 2.10.8-4 First Impact Maximum Inertia g-Load versus Initial Angle of Impact for 30 Foot Drop, Using Minimum Wood Crush Stress Properties .................................................................. 2.10.8-36 Table 2.10.8-5 Maximum Impact Limiter Deformation versus Initial Angle of Impact, for 30 Foot Drop, using the Maximum Wood Properties ............................................................................. 2.10.8-37 Table 2.10.8-6 Maximum Impact Limiter Deformation versus Initial Angle of Impact, for 30 Foot Drop, using the Minimum Wood Properties ............................................................................. 2.10.8-38 Table 2.10.8-7 Maximum Inertial g-Load during 1 Foot Drop........................ 2.10.8-39 Table 2.10.8-8 Maximum Impact Limiter Deformation versus Initial Angle of Impact for 1 Foot Drop ...................................................... 2.10.8-40 Table 2.10.8-9 Loading Used in Cask Body Analysis, Appendix 2.10.1 versus Maximum g-Loads Predicted by ADOC Program ...... 2.10.8-41 Table 2.10.8-10 Loading Used in Basket Structural Analysis, Appendix 2.10.5 versus Maximum g-Load Predicted by ADOC Program ................................................................................ 2.10.8-42 Page 2.10.8-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 2.10.8-1 Impact Limiter Geometry....................................................... 2.10.8-43 Figure 2.10.8-2 Sample Force/Deflection Curve for Balsa ............................. 2.10.8-44 Figure 2.10.8-3 Sample Force/Deflection Curve for Redwood ....................... 2.10.8-45 Figure 2.10.8-4 ADOC Computer Model for TN-40 Transport Package ......... 2.10.8-46 Figure 2.10.8-5 TN-40 Package Geometry during Impact for Wood Strain Computation ......................................................................... 2.10.8-47 Figure 2.10.8-6 Geometry of Packaging ........................................................ 2.10.8-48 Figure 2.10.8-7 Packaging at Time, t ............................................................. 2.10.8-49 Figure 2.10.8-8 Geometry of Impact Limiter Parameters ............................... 2.10.8-50 Figure 2.10.8-9 Definition of Limiter Deformation ........................................... 2.10.8-51 Figure 2.10.8-10 Crush Pattern in Impact Limiter ............................................ 2.10.8-52 Figure 2.10.8-11 Impact Limiter Segments ...................................................... 2.10.8-53 Figure 2.10.8-12 Strain Computation for Crush Pattern I ................................. 2.10.8-54 Figure 2.10.8-13 Strain Computation for Crush Pattern II ................................ 2.10.8-55 Figure 2.10.8-14 Strain Computation for Crush Pattern III ............................... 2.10.8-56 Figure 2.10.8-15 Wood Stress-Strain Curve ...................................................... 2.10.8-57 Figure 2.10.8-16 Impact Limiter Free Body Diagram during 20 Slap Down .... 2.10.8-58 Figure 2.10.8-17 Impact limiter Lifting Lug Geometry ...................................... 2.10.8-59 Figure 2.10.8-18 Cask Geometry during Tip-Over Event ................................. 2.10.8-60 Page 2.10.8-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 STRUCTURAL EVALUATION OF THE IMPACT LIMITERS 2.10.8.1 Introduction This appendix presents the details of the structural analysis of the TN-40 impact limiters. The impact limiters are designed to absorb the kinetic energy resulting from the one (1) foot and thirty (30) foot normal condition of transport (NCT) and hypothetical accident condition (HAC) free drop events specified by 10 CFR 71 [5]. Redwood and balsa wood are used as the primary energy absorption material(s) in the impact limiters.

A sketch of the impact limiter is shown in Figure 2.10.8-1. A functional description of the impact limiters is given in Section 2.10.8.2. The impact limiter design criteria are described in Section 2.10.8.3.

A computer model of the TN-40 Packaging was developed to perform system dynamic analyses during impacts of 30 foot HAC and 1 foot NCT drops. The model was developed for use with the ADOC (Acceleration Due To Drop On Covers) [6] computer code described in detail in Section 2.10.8.8 which determines the deformation of the impact limiters, the forces on the cask and the cask deceleration due to impact of the package on an unyielding surface. Numerous cases were run to determine the effects of the wood properties and the impact angle. A description of the computer model, input data, analysis results and conclusions for the 30 foot HAC condition and one foot NCT free drops are given in Sections 2.10.8.4 and 2.10.8.5 respectively. The analysis of the impact limiter attachments is described in Section 2.10.8.6. A summary of results for all drop orientations is provided in Section 2.10.8.7. The forces and accelerations used in the cask body and basket structural analysis are presented in detail in Appendix 2.10.1 and Appendix 2.10.5 respectively. The accelerations are given in Table 2.10.8-9 and Table 2.10.8-10 (loading values calculated in this appendix are increased for conservatism). Planned testing programs on the TN-40 wood-filled limiters are discussed in Appendix 2.10.9.

2.10.8.2 Design Description The impact limiters absorb energy during impact events by crushing of balsa and redwood. The size, location and orientation of each wood block is selected to provide protection for the cask during all NCT and HAC drop events.

Page 2.10.8-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The top and bottom impact limiters are nearly identical. Each has an outside diameter of 144 inches and a height of 50 inches. The bottom impact limiter has pockets to accommodate the lower trunnions. The inner and outer shells are SA-240 Type 304 stainless steel joined by radial gussets of the same material. The gussets limit the stresses in and deflection of the 0.25 in. thick steel outer cylinder and end plates due to pressure differentials caused by elevation and temperature changes during normal transport and provide wood confinement during impact. The metal structure positions, supports, confines and protects the wood energy absorption material. The metal structure contributes to the energy absorbing capability of the impact limiter. However, the contribution to a side drop or oblique angles is negligible because contact starts at a single point with the unyielding surface (target) and initiates buckling of a single gusset.

After the drop event is complete, relatively few gussets are buckled. The strength of the steel shell is conservatively omitted from the impact limiter analysis.

The region of the impact limiter which is adjacent to the cask ends is filled with balsa wood and redwood mostly oriented with the grain direction parallel to the axis of the cask as shown in Figure 2.10.8-1. The materials and grain orientations are selected to provide acceptably low deceleration to limit stresses in the cask during the 30 foot HAC impact end drop. A 2.50 inch layer of balsa wood with the grain perpendicular to the axis of the cask is provided on the outer face of the impact limiter to minimize decelerations after a one foot NCT end drop.

A 8.75 inch wide ring of redwood (consisting of 12 segments or blocks of wood) is located in the sides of the pie shaped compartments which surround the end of the cylindrical surface of the cask with the grain direction oriented radially. This ring of redwood absorbs most of the kinetic energy during a side drop. Redwood was selected for this portion of the impact limiter because of its high crush strength and hence the ability of a small amount of wood to absorb a large amount of energy in a relatively short crush distance.

The corners of the pie-shaped compartments are filled with redwood. A 34.75 inch section of redwood is located around the outer corner of the impact limiter. The grain is oriented radially. The primary function of the redwood block is energy absorption during a 30 foot corner drop.

All wood blocks used in the impact limiters are composed of individual boards glued together with a Phenol Resorcinol Adhesive or equivalent. This adhesive is selected for its superior strength and moisture resistance. The wood blocks are assembled and glued together in accordance with an approved QA procedure. Minimum properties of the adhesive are listed in Table 2.10.8-1. Ranges of shear and tensile strengths of each type of wood are also listed. The adhesive is significantly stronger than any of the wood used in the limiter in terms of shear and tensile strength. Therefore the boards or blocks of wood will not fail along the glue joints.

Page 2.10.8-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The other mechanical properties of the wood used in the analysis are shown in Table 2.10.8-2. The crush strength properties used cover the range of density and moisture content specified in the fabrication specification. During fabrication, wood samples are tested for density and moisture content in accordance with an approved sampling plan.

If the density or moisture content are not within the specified range, the wood blocks from which samples are taken are rejected.

During the end drop, all of the wood in the central part of the impact limiter that is directly backed-up by the cask body will crush. The wood in the corners and sides of the limiter will tend to slide along the side of the cask since it is not supported or backed-up by the body and it will not crush or absorb energy as effectively as the wood that is backed-up. During the side or oblique drop the wood backed up by the cask will crush, while the wood beyond the end of the cask body will have a tendency to slide around the end of the cask. The analyses assume that the effectiveness of the portion of the wood that is not backed-up is 20%. Effectiveness is defined as the actual crush force developed at the target by this material divided by the theoretical force required to crush the material. The analysis also assumes a range of wood crush strengths. When determining maximum deceleration, the maximum crush strengths are used. When determining crush depth, the minimum wood crush strengths are used.

Each impact limiter is attached to the cask using four attachment bolts and the two limiters are attached to each other by thirteen tierods. The attachments have been sized to withstand the loads transmitted during a low angle drop slap down. This analysis is described in Section 2.10.8.6 of this Appendix.

2.10.8.3 Design Criteria The outside dimensions of the impact limiter are sized to be within federal and state highway height and width restrictions. The balsa and redwood distribution and densities have been selected to limit the maximum cask body inertia loads due to the one foot NCT drop and the thirty foot HAC drop so that the design criteria specified for the cask and basket (Section 2.1) are met.

The welded stainless steel structure of the impact limiter is designed so that the wood is maintained in position and is confined during crushing of the impact limiters. The outer shell and gussets are designed to buckle and crush during impact. Local failure of the shell is allowed during impact limiter crushing. The welded stainless steel shell and its internal gussets are designed to withstand pressure differences and normal handling and transport loads with stresses limited to the material yield strength.

The impact limiters are designed to remain attached to the cask body during all NCT and HAC drop events.

Page 2.10.8-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.8.4 Analysis of the HAC 30 Foot Free Drop 2.10.8.4.1 Approach The kinetic energy due to the HAC 30 foot drop is absorbed by crushing of the impact limiters mounted on the ends of the cask. The limiters contain materials, i.e. balsa and redwood, which provide controlled deceleration of the packaging by crushing between the target surface and the cask body.

The applicable regulation, 10CFR71.73 [5], requires that the packaging be oriented for the drop so that it strikes the target in a position for which maximum damage is expected. Dynamic impact analyses were performed for different packaging orientations using the ADOC computer code described in Section 2.10.8.8. This computer code has been validated by comparing its dynamic results with those from hand calculations for relatively simple problems, comparing its calculated force-deflection curves with those obtained from static crush tests, and by correlating dynamic results with actual measured cask behavior on other programs.

2.10.8.4.2 Assumptions and Boundary Conditions The assumptions and boundary conditions are as follows:

1) The cask body is assumed to be rigid and absorbs no energy. This assumption is realistic since the design criteria of Section 2.1.2 limit metal deformations to small values. All of the impact energy is therefore assumed to be absorbed by the impact limiters.

2) The crushable material is one of several anisotropic materials. The different wood regions are modeled individually.

3) The crush strengths of the wood sections are obtained from the properties parallel to and perpendicular to the grain based on the orientation of the cask at impact.

4) Each wood region is modeled as a one dimensional elastic, perfectly plastic material up to a specific locking strain. After reaching the locking strain, the stress increases linearly with the additional strain. The wood properties (modulus of elasticity, average crush strength, locking modulus, and locking strain) are taken from force-deflection curves of sample blocks of wood. Typical force-deflection curves for balsa and redwood are shown in Figure 2.10.8-2 and Figure 2.10.8-3. Since the locking strain varies from sample to sample, conservatively low locking strains of 80% for balsa and 60% for redwood are used.

5) The crush properties of the wood are based on the initial angle of impact and do not change during the drop event being evaluated.

6) The cask and impact limiters are axisymmetric bodies.

7) The crushing resistance of the impact limiter shell and gussets have a negligible effect on the crush strength of the limiter and, therefore, a negligible effect on the impact forces and inertia loads.

Page 2.10.8-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.8.4.3 Packaging Dynamic Computer Model Figure 2.10.8-4 illustrates the computer model used for all packaging orientations.

Regions I, II, and III in the model are used to delineate regions where different impact limiter materials are used. It should be noted that the properties of the three regions have been designed by choosing wood types and orientations to accommodate the crush requirements of the drop orientations. The crushable materials of Regions I, II, and III are selected to control the decelerations resulting from end, corner, and side drop orientations, respectively. Table 2.10.8-2 tabulates the wood properties that were used to describe the wood stress-strain behavior in the analysis.

A portion of the impact limiter crushable material is backed up by the cask body as it crushes against the impact surface. The remaining material overhangs the cask body and is not backed up. Backed up regions project vertically from the target footprint to the cask body, while unbacked regions do not project vertically to the cask. The effectiveness of the energy absorbing crushable material varies depending on whether it is backed up by the cask or is unsupported. Two cases are analyzed to bound impact limiter performance. In one case, the non-backed up material is assumed to be 20%

effective and maximum wood crush strength is used (maximum of the possible range based on specified density). In the other case, the non-backed up material is also assumed to be 20% effective but the minimum wood strength is used. Evaluating impact limiter performance in this way results in a range of deceleration values, crush forces and crush depths. This, in combination with close control of wood properties during procurement, assures that the effects of wood property variations (including temperature effects) are bounded by the analyses.

2.10.8.4.4 Analysis Results Predicted by ADOC The peak inertia loadings or cask body decelerations (in terms of gs) versus initial angle of impact are presented in Table 2.10.8-3 and Table 2.10.8-4 for the 30 foot drop.

The 30 foot dimension is defined to be the distance from the impact surface to the lowest point of the impact limiter. The center of gravity (CG) of the cask is thus much higher than 30 feet. The values of maximum crush depth for each 30 foot drop orientation are shown in Table 2.10.8-5 and Table 2.10.8-6. Since the TN-40 package CG is within a few inches of the geometric package center and the impact limiters are nearly identical, these tables are valid for impacts on either end.

2.10.8.4.5 Wood Strain Computation During the low angle drops (5° to 40°) the method used by ADOC to compute the strain in the wood segments is overly conservative. Consequently, the maximum strains achieved during certain impacts with drop orientations between 5 and 40 are recomputed. The average wood strains are only recomputed for the minimum wood property cases since strains generated using minimum wood properties bounds the maximum wood property strain for a given drop orientation. The Figure 2.10.8-5 depicts the geometry of the package during impact.

Page 2.10.8-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Given the impact angle and the crush depth , computed by ADOC, the crush depths a and b shown in Figure 2.10.8-5 can be calculated. The impact limiter strain is then taken to be the average of a and b divided by the impact limiter radial thickness, 25.50 inches. The recomputed strains are provided in Table 2.10.8-5 and Table 2.10.8-6, identified by *.

2.10.8.4.6 Trunnion Clearance during Low Angle Drops In this section the clearance between the impact target and the outer edge of the trunnions is computed in order to assure that the trunnions are not damaged or cause large accelerations during a near 0° drop. The trunnion clearance is only computed for the minimum wood property case since crush depths are always smaller when maximum wood properties are used.

Trunnion Clearance during 0° Side Drop The maximum crush depth sd during a side drop event (using minimum wood properties) is:

sd = 15.87 in.

The trunnion clearance sd is then:

sd = 71.75 in. (impact limiter outer radius) - 15.87 in. - 104.50 / 2 in. (trunnion radius)

= 3.63 in.

Trunnion Clearance during 5° Drop The maximum crush depth 5° during a side drop event (using minimum wood properties) is:

5° = 18.89 in.

The angle of the cask at the time of the maximum crush depth, 5° is:

5° = 90° - 88.61° = 1.39° The trunnion clearance 5° is then:

5° = 71.75 in. (impact limiter outer radius) - [ 18.89 in. - 37.25 sin(1.39°) ]

- 104.50 / 2 in. (trunnion radius) = 1.51 in.

Trunnion Clearance during 10° Drop The maximum crush depth 10° during a side drop event (using minimum wood properties) is:

10° = 18.15 in.

Page 2.10.8-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The angle of the cask at the time of the maximum crush depth, 10° is:

10° = 90° - 88.24° = 1.76° The trunnion clearance 10° is then, 10° = 71.75 in. (impact limiter outer radius) - [ 18.15 in. - 37.25 sin(1.76°) ]

- 104.50 / 2 in. (trunnion radius) = 2.49 in.

Trunnion Clearance during 15° Drop The maximum crush depth 15° during a side drop event (using minimum wood properties) is:

15° = 20.06 in.

The angle of the cask at the time of the maximum crush depth, 15° is:

15° = 90° - 76.79° = 13.21° The trunnion clearance 15° is then:

15° = 71.75 in. (impact limiter outer radius) - [ 20.06 in. - 37.25 sin(13.21°) ]

- 104.50 / 2 in. (trunnion radius) = 7.95 in.

Based on the above calculations, there is adequate clearance between the trunnion outer edge and the impact target during low angle 30 foot drops.

2.10.8.5 Analysis for One Foot Drop Normal Condition of Transport This section describes the analysis of the TN-40 packaging for the one foot NCT drop.

The TN-40 cask is lifted vertically and is transported horizontally. End and side drop orientations are therefore considered to be credible NCT drop events. Any other drop orientation will cause the cask to tip over onto its side, which is clearly an accident. The accident analyses in Section 2.10.8.4 bound any possible tipping accident. Therefore, the one foot drop analysis is performed for end and side drop orientations. A one foot, 63.8 CG over corner drop is also analyzed to show that the NCT side and end drops are critical with respect to acceleration and deformation.

The packaging kinetic energy is absorbed by crushing of the impact limiters. The dynamic system model of Section 2.10.8.4 was used to perform the side drop (0) analysis using the ADOC computer program described in Section 2.10.8.8. The end drop analysis was performed assuming that the energy would be absorbed by the soft balsa wood (oriented in the weak direction) in the outer end of the limiter. This is an accurate way to determine g loads on an end drop since the g values can be calculated by the expression F = Ma where F is the crush stress times the area and M is the package weight divided by the acceleration of gravity g.

Page 2.10.8-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The inertial load results of these one foot drop analyses are presented in Table 2.10.8-7. Again, two extreme cases are considered. The upper bound stiffness case assumes maximum wood crush strength and the lower bound stiffness case assumes minimum wood strength. Stress analyses in Section 2.10.1 are conservatively performed for the case(s) with maximum inertia loads resulting from upper bound stiffness cases.

The maximum crush depths for each of the 1 foot drop orientations are presented in Table 2.10.8-8.

2.10.8.6 Impact Limiter Attachment Analysis The impact limiter attachments are designed to keep the impact limiters attached to the cask body during all NCT and HAC events. The loading that has the highest potential for detaching the impact limiter is the slap down or secondary impact after a shallow angle 30 foot drop. During this impact, the crushing force on the portion of the impact limiter beyond the cask body (the non backed-up area) tends to pull the limiter away from the cask. The end and corner drops are not critical cases for the impact limiter attachments since the impact force tends to push the impact limiter onto the cask in these orientations.

2.10.8.6.1 Tie Rod Stress Analysis For the tie rod evaluation, maximum wood crush strengths of 2010 psi for balsa and 6500 psi for redwood are assumed. The maximum wood properties produce the highest overturning moment on the limiter. Based on the dynamic analysis performed using the ADOC code, the most severe slap down impact occurs after a shallow angle oblique impact.

The worst case loading applied to the impact limiter attachments occurs during a low angle (20°) slap down event. The maximum lateral force due to 20 second (slap down) impact is 12,500 kips at a distance of 82.3 inches from the package C.G. This load is conservatively set higher than the actual force of 12,210 kips computed by the ADOC analysis. Only the thirteen 1.5 inch diameter tie rods react the moment applied during the 20 slap down drop. This assumption is considered conservative because the impact limiter attachment bolts will also take some of this load.

The maximum moment applied to the impact limiter attachments is conservatively determined ignoring the mass of the impact limiter which tends to reduce the attachment forces. A free body diagram of the impact limiter is shown in Figure 2.10.8-16. It is conservatively assumed that the impact limiter pivots about the edge of the cask. The resultant of the external impact force on the limiter is offset 5.16 in. from the resultant of the cask reaction force. Therefore, the net moment applied to the limiter is 12.50106 5.16 or 6.45 107 in lb in the counterclockwise direction. There is also a frictional force that acts to pull the impact limiter away from the cask. Assuming a frictional coefficient of 0.42 between the cask and limiter and between the limiter and impact surface, the magnitude of this force is:

Page 2.10.8-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Ff = R = (0.42)(1.25107) = 5.25106 lbs.

The crush depth on the side is 5.62 inches. The resultant moment due to friction is:

Mf = (5.25106)(26.50 - 5.62) = 1.0962108 in lbs. (clockwise)

The total moment is therefore 4.512107 in lbs in the clockwise direction. Assume that only the tie rods hold the impact limiters in place, and that the impact limiter will tend to pivot around the edge of the cask. The force distribution among the tie rods will be linearly proportional to their distance from the pivot point. There are two different angular orientations of the impact limiter that are of interest. The first orientation is the angle that causes the highest stress in the single tie rod brackets, and the second is the angle that causes the highest stress in the double tie rod brackets.

Orientation 1 Stress Analysis During a slapdown impact event in orientation 1, the target surface crushes the impact limiters at the double tie rod bracket located at 270° (Appendix 1.4.1 of Chapter 1, Drawing 10421-71-40). The angular location of each tie rod bracket from vertical (perpendicular to the target surface) is computed in the following way. The tie rods are numbered clockwise 1 through 13 starting from the first tie rod located just above the 270° mark.

Tie Rod Tie Rod Angular Location Calculation Angular Location Number Bracket Type (For Orientation 1) 1 Double 180° + arctan[4.31/(124.00/2)] = 184° 2 Single 180° + 30° = 210° 3 Double 180° + 90° - arctan[4.31/63.00] = 266° 4 Double 180° + 90° + arctan[4.31/63.00] = 274° 5 Single 180° + 120° = 300° 6 Single 180° + 150° = 330° 7 Single 180° + 180° = 360° 8 Single 180° + 210° = 390° 9 Single 180° + 240° = 420° 10 Double 180° + 270° - arctan[4.31/63.00] = 446° 11 Double 180° + 270° + arctan[4.31/63.00] = 454° 12 Single 180° + 330° = 510° 13 Double 360°+180° - arctan[4.31/(124.00/2)] = 536° Page 2.10.8-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The vertical distance from the target surface to each tie rod is computed as follows.

Tie Rod Tie Rod Vertical Distance Calculation (in.) Vertical Number Bracket Type (For Orientation 1) Distance (in.)

1 Double [4.312 + (124/2)2]1/2 cos(184°) + 91.00/2 = -16.50 2 Single (124.00/2) cos(210°) + 91.00/2 = -8.19 3 Double [4.312 + 63.002]1/2 cos(266°) + 91.00/2 = 41.09 4 Double [4.312 + 63.002]1/2 cos(274°) + 91.00/2 = 49.90 5 Single (124.00/2) cos(300°) + 91.00/2 = 76.50 6 Single (124.00/2) cos(330°) + 91.00/2 = 99.19 7 Single (124.00/2) cos(360°) + 91.00/2 = 107.50 8 Single (124.00/2) cos(390°) + 91.00/2 = 99.19 9 Single (124.00/2) cos(420°) + 91.00/2 = 76.50 10 Double [4.312 + 63.002]1/2 cos(446°) + 91.00/2 = 49.90 11 Double [4.312 + 63.002]1/2 cos(454°) + 91.00/2 = 41.09 12 Single (124.00/2) cos(510°) + 91.00/2 = -8.19 13 Double [4.312 + (124/2)2]1/2 cos(536°) + 91.00/2 = -16.50 In the table above, negative vertical distances correspond to tie rods that are below the crush line of the impact limiter. Therefore, these tie rods are conservatively considered to be ineffective and do not carry any of the prying load.

The tie rod farthest away from the target surface (tie rod 7) is assumed to carry the maximum tensile force, Fmax. All other tie rods are assumed to carry a tensile force linearly proportional to their distance from the target surface. Therefore, each tie rod will carry the following prying moment.

Tie Rod Tie Rod Tensile Force Moment Arm Moment Number Bracket Type in Orientation 1 (lb.) (in.) (in.lb.)

1 Double 0 - 0 2 Single 0 - 0 3 Double (41.09/107.50) Fmax 41.09 15.71 Fmax 4 Double (49.90/107.50) Fmax 49.90 23.16 Fmax 5 Single (76.50/107.50) Fmax 76.50 54.44 Fmax 6 Single (99.19/107.50) Fmax 99.19 91.53 Fmax 7 Single Fmax 107.50 107.50 Fmax 8 Single (99.19/107.50) Fmax 99.19 91.53 Fmax 9 Single (76.50/107.50) Fmax 76.50 54.44 Fmax 10 Double (49.90/107.50) Fmax 49.90 23.16 Fmax 11 Double (41.09/107.50) Fmax 41.09 15.71 Fmax 12 Single 0 - 0 13 Double 0 - 0 Total Moment = 477.18 Fmax Page 2.10.8-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Equating this total moment with Mtot computed above and solve for Fmax:

477.18 Fmax = 4.512107 in.lb. Fmax = 4.512107 / 477.18 = 94,556 lb.

The minimum tensile area of the 11/2 inch diameter tie rod is at the threads and is 1.490 in2. The maximum allowable stress is lesser of 0.7Su or Sy (Level D, Bolted joint, F1335.1, [4]). The maximum impact limiter surface temperature at bracket attach to tie-rod is around 120°F (Chapter 3, Figure 3-10), conservatively 200°F is used for the material allowable. The tie rod material is A193 Grade B7, which has an ultimate strength, Su, of 125 ksi [4] at 200°F and a yield strength of Sy = 98 ksi [4] at 200°F.

Therefore, the allowable is smaller of 0.7 x 125 ksi = 87.5 ksi or Sy = 98 ksi. So for orientation 1, the maximum tie rod tensile stress is 94,556 / 1.490 = 63.5 ksi, which is less than 0.7 Su = 87.5 ksi.

Orientation 2 Stress Analysis During a slapdown impact event in orientation 2, the target surface crushes the impact limiters at the double tie rod bracket located at 0° (Appendix 1.4, Drawing 10421-71-40).

The angular locations of each tie rod bracket from vertical (perpendicular to the target surface) are computed in the following way. The tie rods are numbered clockwise 1 through 13 starting from the first tie rod located just above the 270° mark.

Tie Rod Tie Rod Angular Location Calculation Angular Number Bracket Type (For Orientation 2) Location 1 Double 90° + arctan[4.31/(124.00/2)] = 94° 2 Single 90° + 30° = 120° 3 Double 90° + 90° - arctan[4.31/63.00] = 176° 4 Double 90° + 90° + arctan[4.31/63.00] = 184° 5 Single 90° + 120° = 210° 6 Single 90° + 150° = 240° 7 Single 90° + 180° = 270° 8 Single 90° + 210° = 300° 9 Single 90° + 240° = 330° 10 Double 90° + 270° - arctan[4.31/63.00] = 356° 11 Double 90° + 270° + arctan[4.31/63.00] = 364° 12 Single 90° + 330° = 420° 13 Double 360°+ 90° - arctan[4.31/(124.00/2)] = 446° Page 2.10.8-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Compute the vertical distance from the target surface to each tie rod as follows.

Tie Rod Tie Rod Vertical Distance Calculation (in.) Vertical Number Bracket Type (For Orientation 2) Distance (in.)

1 Double [4.312 + (124/2)2]1/2 cos(94°) + 91.00/2 = 41.16 2 Single (124.00/2) cos(120°) + 91.00/2 = 14.50 3 Double [4.312 + 63.002]1/2 cos(176°) + 91.00/2 = -17.50 4 Double [4.312 + 63.002]1/2 cos(184°) + 91.00/2 = -17.50 5 Single (124.00/2) cos(210°) + 91.00/2 = -8.19 6 Single (124.00/2) cos(240°) + 91.00/2 = 14.50 7 Single (124.00/2) cos(270°) + 91.00/2 = 45.50 8 Single (124.00/2) cos(300°) + 91.00/2 = 76.50 9 Single (124.00/2) cos(330°) + 91.00/2 = 99.19 10 Double [4.312 + 63.002]1/2 cos(356°) + 91.00/2 = 108.49 11 Double [4.312 + 63.002]1/2 cos(364°) + 91.00/2 = 108.50 12 Single (124.00/2) cos(420°) + 91.00/2 = 76.50 13 Double [4.312 + (124/2)2]1/2 cos(446°) + 91.00/2 = 49.84 In the table above, negative vertical distances correspond to tie rods that are below the crush line of the impact limiter. Therefore, these tie rods are conservatively considered to be ineffective and not carry any of the prying load.

The set of tie rods farthest away from the target surface (tie rods 10 and 11) are assumed to carry the maximum tensile force, Fmax. All other tie rods are assumed to carry a tensile force linearly proportional to their distance from the target surface.

Therefore, each tie rod will carry the following prying moment.

Tensile Force Tie Rod Tie Rod Moment Arm Moment In Orientation 2 Number Bracket Type (in.) (in.lb.)

(lb.)

1 Double (41.16/108.50) Fmax 41.16 15.61 Fmax 2 Single (14.50/108.50) Fmax 14.50 1.94 Fmax 3 Double 0 - 0 4 Double 0 - 0 5 Single 0 - 0 6 Single (14.50/108.50) Fmax 14.50 1.94 Fmax 7 Single (45.50/108.50) Fmax 45.50 19.08 Fmax 8 Single (76.50/108.50) Fmax 76.50 53.94 Fmax 9 Single (99.19/108.50) Fmax 99.19 90.68 Fmax 10 Double Fmax 108.50 108.50 Fmax 11 Double Fmax 108.50 108.49 Fmax 12 Single (76.50/108.50) Fmax 76.50 53.94 Fmax 13 Double (49.81/108.50) Fmax 49.81 22.89 Fmax Total Moment = 477.01 Fmax Page 2.10.8-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Equating this total moment with Mtot computed above and solve for Fmax:

477.01 Fmax = 4.512107 in.lb. Fmax = 4.512107 / 477.01 = 94,589 lb.

The minimum tensile area of the 11/2 inch diameter tie rod is at the threads and is 1.490 in2. So for orientation 2, the maximum tie rod tensile stress is 94,589. / 1.490 = 63.5 ksi, which is less than 0.7 Su = 87.5 ksi.

2.10.8.6.2 Tie Rod Bracket Analysis Tie Rod Bracket And Impact Limiter Gusset Allowable Stress The material used for the tie rod brackets and impact limiter gussets is A-240 Type 304.

The allowable stress for the brackets and impact limiter gussets is Su or 71 ksi [4] at 200°F.

Tie Rod Bracket / Impact Limiter Weld For 1/4 inch fillet welds, the throat width is 0.25 sin (45°) = 0.1768 in.

Single bracket:

Area of weld = 0.1768 x 2(13 + 6) = 6.718 in.2 weld = 94,589 / 6.718 = 14,080 psi < 71,000 psi Double bracket:

Area of weld = 0.17682(11 + 21.62) = 11.534 in.2 weld = 294,589 / 11.534 = 16,402 psi < 71,000 psi Impact limiter gussets Since the gussets are fillet welded on both sides to the top plate of the impact limiter, the cross sectional area of the gusset is the critical tensile area. Assume the tensile force from the tie rods acts over a length of the gusset equal to 150% of the length of the bracket reinforcement pad.

Single bracket:

Active tensile area of gusset = (0.19 in.)(1.56.00 in.) = 1.710 in.2 gusset = 94,589 / 1.710 = 55,315 psi < 71,000 psi Double bracket:

Active tensile area of gusset = (0.19 in.)(1.511.00 in.) = 3.135 in.2 gusset = 2 x 94,589 / 3.135 = 60,344 psi < 71,000 psi Page 2.10.8-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Fillet weld / side plate Single bracket:

Throat width = 2 x 0.31 (45°) = 0.438 in.

Stress Area = 520.438 = 4.384 in.2 weld = 94,589 / 4.384 = 21,576 psi < 71,000 psi Double bracket:

Throat width = 2 x 0.31 sin (45°) = 0.438 in.

Stress Area = 2(2)(.25) sin (45°) + 2(5)(.438) = 5.087 in.2 weld = 294,589 / 5.087 = 37,189 psi < 71,000 psi Shear stress in 2 inch top plate (allowable shear stress 0.42 Su)

Single bracket:

Shear Area = 2(7.00 in. - 1.75 in.) = 10.5 in.2

= 94,589 / 10.5 = 9,008 psi < 29,820 psi Double bracket:

Shear Area = 2 (7.00 in. - 1.75 in.) = 10.5 in.2

= 2 x 94,589 / 10.5 = 18,017 psi < 29,820 psi 2.10.8.6.3 Lifting Lug Analysis The weights of the top and bottom impact limiters are 16,338 lb and 16,332 lb respectively. For the following analysis conservatively use a weight of 16,450 lb. per impact limiter. Each impact limiter is supported by two lifting lugs. The material used for the lifting lugs is also A-240 Type 304. The temperature at outer surface of impact limiter is around 120°F (Chapter 3, Figure 3-10), conservatively 150°F is used for the lifting lug analysis. The allowable shear stress is 0.5 Sy = 13.75 ksi. The allowable primary plus bending is Sy = 27.5 ksi [4].

A 60 angle between the slings and horizontal is assumed. This allows the slings to clear the impact limiter. The loading is taken to be 3 times the weight of the impact limiter. The lifting lugs are on the impact limiter CG. Figure 2.10.8-17 shows the geometry of the impact limiter lifting lugs.

The tension in the lifting sling is, 2 16,450 1 T= = 28,492lbs 2 sin( 60 )

Page 2.10.8-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The normal stress n in the lifting lug is conservatively computed in the following way:

Normal Force Fn = 28,492 sin(30°) = 14,246 lb Normal Area, A = (1)(5-1.5) = 3.5 in.2 Normal Stress n = 14,246/3.5 = 4,070 psi The bending stress b in the lifting lug is computed in the following way.

Moment of Inertia, I = (1/12)(1)(53) = 10.42 in.4 Fb = 28,492 (cos (30°)) = 24,675 lb 2.5(2.5)(24,675 )

b = = 14,800 psi 10 .42 The shear stress in the lug is conservatively computed as follows.

Shear Area = 2x(2.5 - 0.75)(1) = 3.50 in.2

= 28,492/3.50 = 8,141 psi.

The total stress intensity is,

( 4,070 + 14,800 )2 + 4(8,141)2 S.I. = = 24,924 psi. < 27,500 psi.

The stresses in the lifting lug weld are computed as follows.

Stress Area, Alug = (0.375) sin(45)(52+12) = 3.18 in2 Normal Stress n = 14,246/3.18 = 4,480 psi Moment of inertia, Iweld = (2/12)(0.375)sin(45)(53) + (2/12)(1)[0.375 sin(45)]3

+ 2[0.375 sin(45)](1)(2.52) = 8.84 in4.

2.5( 2.5 )( 24,675 )

b = = 17,446psi 8.84

= 24,675 / 3.18 = 7,759 psi

( 4,480 + 17,446 )2 + 4(7,759 )2 S.I. = = 26,862 psi. < 27,500 psi.

Page 2.10.8-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.8.6.4 Impact Limiter Attachment Bolt and Bracket Analysis The thirteen tie rods attached to both impact limiters are designed to hold the impact limiter on the cask during all drop scenarios without the aid of the attachment bolts. The purpose of the eight attachment bolts and brackets is the hold the top impact limiter on the cask body during a tip over event immediately following a near 90 corner drop.

After a high angle corner drop (45 to 90) the crushing of the impact limiter (from inside where the cask contacts the impact limiter) on the bottom (impact side) could cause the tie rods to become loose. In the event that the tie rods become loose, and the package tips over (second impact) the attachment bolts will hold the top impact limiter in place.

The following calculation shows that the attachment bolts and brackets are structurally adequate to withstand the load corresponding to a tip over (second impact) event.

Four 11/2 - 8UN bolts are used to attach the top impact limiter to the cask in the event of a 30 foot corner drop where the impact limiter crushing exceeds 12.00 inches from the inside where the cask tips over (immediately after a corner drop) when the top impact limiters inertia would tend to pull it off the cask.

During the tip over, after a corner drop, simple energy conservation laws are used to determine the maximum axial acceleration that the top impact limiter could experience.

All of the potential energy of the package in the CG over corner drop impact position is assumed to be converted into rotational energy (the CG over corner drop impact position yields the highest potential energy). This is conservative since some of the potential energy in the package is converted to vertical translational energy. Figure 2.10.8-18 shows the height of the center of gravity of the package in the impact position, H1, and after the tip over, H2.

The axial distance from the bottom of the bottom impact limiter to the CG is given by:

91.42 in. (distance from bottom of cask to CG) + 38.00 in. (thickness of impact limiter)

= 129.42 in.

The weight of the TN-40 Transport Package is W = 271,455 lb. For this analysis, a weight of 272,000 lb is conservatively used. Therefore, the potential energy change of the package during tip over PE is therefore:

PE = W H = 272,000 lb. (148.10 - 72.00) = 2.070107 in. lb.

Where, H is the change in height of the package during tip over. This energy is assumed to be converted entirely into rotational energy of the package KE which is equal to the following:

KE = 1/2 I 2 Where I is the moment of inertia of the package about the pivot point, and is the rate of rotation of the cask after tip over. The moment of inertia of the TN-40 transport package about its center of gravity, ICG, is 3.393106 lbf.in.sec2 [Section 2.2]. Therefore the moment of inertia of the package about the pivot point is:

Page 2.10.8-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 I = ICG + m r2 = 3.393106 + (272,000/386.4) 148.102 = 1.883107 lbf.in.sec2.

Then, PE =KE = 1/2 I 2 2 = 2 PE / I = 2 2.070107 / 1.883 107 = 2.2 rad2 / sec2 The axial centripetal acceleration a associated with this angular velocity is:

a = r 2 = 270.62 2.2 = 595.4 in.sec.-2 Here, r is the distance from the pivot point to the end of the top impact limiter ( r =

[260.872 + 722]1/2 = 270.62 in). The corresponding g-load is a / g = 595.4/386.4 = 1.54 g.

Therefore, for analysis purpose, assume that the top impact limiter experiences 1.6 g during tip over. A conservative weight of 16,500 lb is used for the top impact limiter.

Therefore, the tensile force applied to the four bolts is:

16,5001.6 = 26,400 lbs.

The tensile force per bolt is, 26,400 / 4 = 6,600 lb/bolt Attachment Bolt Stress Analysis The Stress area for a 11/2 - 8UN bolt is 1.4899 in2 [3]. Therefore the tensile stress is 6,600 / 1.4899 = 4,430 psi.

The impact limiter bolt material is A540 Class 2 with allowable stress equal to the smaller of 0.7 Su = 108.5 ksi (Su = 155 ksi at 250°F [4]) or Sy = 131.55 ksi [4] at 250°.

The above calculated stress of 4430 psi is well below the allowable stress of 108.5 ksi.

Attachment Bolt Bracket Stress Analysis The geometry of the TN-40 impact limiter attachment bolt is provided in Drawing 10421-71-44 in Chapter 1, Appendix 1.4.1. The load applied to the bracket by the bolt is counteracted by shear force in the fillet weld between the bracket and the outer shell.

The resulting shear stress in the weld is calculated in the following way. The throat area of the weld At is:

At = 0.75cos(45°) = 0.530 in.

The weld shear area is 7x0.530 = 3.712 in2 (weld length is 7 inches, so that the bolt holes are not welded over).

The shear stress per bracket, , is:

Page 2.10.8-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

= 6,600 / 3.712 = 1,778 psi.

The bracket material is A516 Grade 70. Therefore the allowable shear stress is 0.42 Su or 29.4 ksi (Su = 70 ksi at 250°F [4]), which is well above the calculated shear stress.

The bending stress in the bolting bracket is computed assuming that only the 5 inch tall plate carries the bending load. The moment of inertia of the plate is:

10 x0.75 3 I= = 0.3516 in.4 12 The applied moment is:

M = 6,6002.66 = 17,556 in. lb.

Mc 17,556x(0.75 2)

= = = 18,724 psi. < 70,000 psi.

I 0.3516 2.10.8.6.5 Impact Limiter Shell Stress Analysis Using the geometry of the TN-40 impact limiter bolt bracket and gusset plate/bolting boss from drawings 10421-71-43 and 10421-71-44, in Chapter 1, Appendix 1.4.1, a stress analysis is performed on the impact limiter bolt connection. The material used for the gusset plate is A-240 Type 304 with Sy = 23.75 ksi, and Su = 68.5 ksi [4] at 250°F.

Conservatively assuming the entire load is carried by the gusset plate/welds and that the 1/4 inch thick impact limiter plate does not bear any of the bolt load the corresponding stresses are analyzed. The length of the moment arm acting on the gusset plate and bottom gusset plate weld is measured from the bolting boss center (where the load is applied) to the termination of the gusset plate at the side impact limiter plate which is 7.16 inches. Referring to the bolt load calculated in section 2.10.8.6.4 the applied moment on the gusset plate and attachment weld is:

6,600 lb(7.16 inches)= 47,256 in lb Similarly the moment arm for the gusset plate/bolting boss interface is measured from the center of the bolting boss to the gusset plate/bolting boss connection which results in an applied moment of:

6,600 lb(1.5 inches)= 9,900 in lb Since both welds have the same stress area and moment of inertia, the bottom weld is the limiting case for analysis. The stress area and moment of inertia for the bottom weld are:

Stress Area = 0.25sin(45°)(2(3.5+0.375))= 1.37 in2 Page 2.10.8-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Moment of inertia, Iweld = (2/12)(0.25)sin(45°)(3.53) + (2/12)(0.25)[0.25 sin(45°)]3

+ 2[0.25 sin(45°)](0.25)(3.52) = 1.67 in4 The corresponding shear and bending stresses in the bottom weld are then Shear Stress, = 6,600 / 1.37 = 4,818 psi < 0.42 Su (28,770 psi)

Mc 47 ,256 (3.5 )

2 = 49 ,520 psi < S (68,500 psi)

Bending Stress, b = = u I 1.67 Using a similar analysis for the gusset plate gives a stress area and moment of inertia of Stress Area = 0.375(3.5)= 1.3125 in2 and Moment of inertia, Iplate = (1/12)(0.375)(3.53)= 1.34 in4 The corresponding shear and bending stresses in the gusset plate are then Shear Stress, = 6,600 / 1.3125 = 5,029 psi < 0.42 Su (28,770 psi)

Mc 47,256 (3.5 )

2 = 61,715 psi. < S (68,500 psi)

Bending Stress, b = = u I 1.34 2.10.8.6.6 Attachment Bolt Torque Assume a bolt force, Fa, of 6,600. The torque required is:

Q = K Db Fa = 0.11.56,600 = 990 in.lb. = 82.5 ft.lb.

Where Fa is the bolt force, Q is the applied torque, K is the nut factor (0.1 with lubrication), and Db is the nominal bolt diameter.

For a bolt torque of 60 ft. lb.,

Q 60 12 Fa = = = 4,800 lb.

KD b 0.1 1.5 For a bolt torque of 80 ft. lb.,

Q 80 12 Fa = = = 6,400 lb.

KD b 0.1 1.5 Page 2.10.8-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Therefore, the maximum tensile stress in the bolt due to pretension is 6,400/1.4899 =

4,296 psi which is less than 0.7Su = 108.5 ksi.

2.10.8.7 Summary of ADOC Results Used for Structural Analysis Cask Structural Analysis, g-Load and Drop Orientation In order to determine the cask stresses, the maximum g-loads from ADOC runs are converted to forces and applied as quasistatic loadings on the cask body. A detailed ANSYS finite element model of the TN-40 cask is used to perform this analysis.

Only the loads corresponding to the most critical normal and accident condition free drop orientations are used in the cask body analysis in Appendix 2.10.1. For the 30 foot accident condition drops, g-loads corresponding to four different angles are evaluated, and for the 1 foot normal condition drops, g-loads corresponding to two different angles are evaluated. The orientations evaluated in Appendix 2.10.1 are as follows.

Drop Height Orientation (Normal / Accident) Analyzed 0 Side Drop 30 Foot 20 Slap Down Accident Condition Drop 63.8 C.G. Over Corner Drop 90 End Drop 1 Foot 0 Side Drop Normal Condition Drop 90 End Drop The g-loads corresponding to these drop orientations are provided in Table 2.10.8-3, Table 2.10.8-4, and Table 2.10.8-7.

The thirty foot side drop is evaluated because it produces the highest normal transverse g-load. The 20, thirty foot slap down is analyzed because it produces a high normal as well as rotational g-load at the ends of the cask (second impact). Stresses in the cask and lid bolts are most sensitive to g-loads applied in the 63.8 (CG over corner) direction. Consequently, the thirty foot CG over corner drop is evaluated. The highest axial g-load occurs during a 90, thirty foot end drop, and is therefore also evaluated.

For the normal condition one foot drops, the 0 side drop, and the 90 end drop are bounding, since they produce the highest normal g-loads in the transverse and axial directions respectively. The g-loads from other drop angles are small and generate insignificant rotational inertia g-loads due to much lower impact velocity.

The g-loads predicted by ADOC and used in the cask body analysis are shown in Table 2.10.8-9.

Page 2.10.8-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Basket Structural Analysis, g-Load and Drop Orientation The loading conditions considered in the evaluation of the fuel basket consist of inertial loads resulting from normal handling (1 foot drop) and hypothetical accident (30 foot) drops. The inertial loads of significance for the basket analysis are those that act transverse to the cask and basket structural longitudinal axes, so that the loading from the fuel assemblies is applied normal to the basket plates and is transferred to the cask wall by the basket. For the slapdown drop case, the second impact (combined transverse g load and rotational g load) will have more severe impact to the component than the first impact. These combined g loads are higher than the side drop g loads; therefore, for the basket side drop analyses, the g load used for analyses bounds both side drop and slap down g loads.

Chapter 2, Section 2.7.1 established the baseline g loads (based on the testing results) used for the basket structural analyses.

2.10.8.8 Summary Description of ADOC Computer Code One of the accident conditions which must be evaluated in the design of transport packagings to be used for the shipment of radioactive material is a free drop from a thirty-foot height onto an unyielding surface (10CFR71). The packaging must be dropped at an orientation that results in the most severe damage. Impact limiters are usually provided on the packaging to cushion the effects of such impact on the containment portion of the packaging. The limiters are usually hollow cylindrical cups which encase each end of the containment and are filled with an energy absorbing material such as wood.

A computer code, ADOC (Acceleration due to Drop On Covers), has been written to determine the response of a packaging during impact. The analysis upon which this code is based is discussed in this section. The overall analysis of the packaging response is discussed in Section 2.10.8.8.1, and the methods used to compute the forces in the limiters as they crush are presented in Section 2.10.8.8.2.

2.10.8.8.1 General Formulation The general formulation used to compute the response of the packaging as it impacts with a rigid target is discussed in this section. The assumptions upon which the analysis is based are first presented followed by a detailed development of the equations of motion used to calculate the packaging dynamic behavior. This is followed by a discussion of the numerical methods and the computer code used to implement the analysis. A significant part of the development is concerned with the prediction of forces developed in the impact limiters as the impact occurs. This aspect of the evaluation is discussed in Section 2.10.8.8.2.

Page 2.10.8-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Assumptions The cask body is assumed to be rigid and axisymmetric. Therefore, all of the energy absorption occurs in the impact limiters which are also assumed to have an axisymmetric geometry. Several assumptions are made in calculating the forces which develop in the limiters as they crush. These are discussed in Section 2.10.8.8.2. Since the packaging is axisymmetric, its motion during impact will be planar. The vertical, horizontal, and rotational components of the motion of the packaging center of gravity (CG) are used to describe this planar motion.

Equations of Motion A sketch of the packaging at the moment of impact is shown on Figure 2.10.8-6. The packaging is dropped from a height H, measured from the lowest point on the packaging to the target. The packaging is oriented during the drop, and at impact, so that the centerline is at an angle r with respect to the horizontal. At the instant of impact, the packaging has a vertical velocity of V0 = 2gH

. (1)

Where g is the gravitational constant.

At some time t after first impact, the packaging has undergone vertical u, horizontal x, and rotational displacements. The location of the packaging at this time is shown on Figure 2.10.8-7. One or both of the limiters have been crushed as shown. The resulting deformations (and strains) in the limiters result in forces which the limiters exert on the packaging, thereby decelerating it. These forces, and their points of application on the packaging, are shown on Figure 2.10.8-7 as Fv1, Fv2, and Fh. The method used to calculate these forces and the points of application are provided in Section 2.10.8.8.2, below.

The three equations of motion of the cask are:

Mu + Fv 1 + Fv 2 W = 0 , (2)

Mx Fh = 0 , and (3)

J Fv 1xv 1 + Fv 2 xv 2 + FhYh = 0 . (4)

Where M is the mass of packaging, J is the polar moment of inertia of the packaging about its CG, W is the packaging weight, and denotes acceleration. At impact (t = 0),

all of the initial conditions are zero except that u = the vertical velocity.

Page 2.10.8-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Computer Solution The computer code is written to compute the motion of the packaging during impact.

The solution is obtained by numerically integrating the equations of motion (equations 2, 3, and 4) from the time of impact, t = 0, to a specified maximum time, tmax. The integrations are carried forward in time at a specified time increment, t. Parametric studies indicate that a time increment of 1 msec is sufficiently small so that further reduction of the time increment does not affect the results. Solutions are usually carried out to about 150 msec for the near horizontal drops and to about 50 msec for the near vertical drops. The significant motions of the packaging normally occur within these time periods.

A standard fourth order Runge Kutta numerical integration method is used to perform the numerical integrations. The following procedure is used to carry the solution from time ti to time ti+1. Note that at time ti the displacements and velocities of the three degrees of freedom describing the motion of the CG of the packaging are known.

1) Calculate the deformation of each of the limiters based on the packaging geometry and the motion of the package CG (see Section 2.10.8.8.2).

2) Calculate the forces which the limiters exert on the packaging body using the deformation of the limiters and their stress-strain characteristics (see Section 2.10.8.8.2).

3) Use Equations 2, 3, and 4 to calculate the accelerations during the time interval. Use the Runge Kutta equations to calculate the location and velocity of the cask CG at time ti+1.

4) Go to step (1) to repeat the process until time tmax.

5) Generate the output.

Output from the code consists of:

Problem title, packaging geometry, drop conditions, and integration data.

Limiter geometric and material property data.

History of packaging CG motion and amount of crushing in each of the limiters.

Force history data.

Plot of acceleration histories.

Plot of maximum limiter deformations.

Page 2.10.8-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.8.8.2 Forces in Limiters The methods used to calculate the forces Fv1, Fv2, and Fh in the limiter at a given crush depth are discussed in this section. These calculations are used to perform steps (1) and (2) above. The limiter geometry and material specification is discussed first. The general methodology used to calculate the forces are then presented which is followed with a detailed development of the equations used to calculate the force-displacement relationships.

Limiter Geometry A sketch of the model of a limiter is shown on Figure 2.10.8-8. Regions I, II and III are used to delineate regions where different materials are used. It should be noted that the properties of the three regions are designed to accommodate the crush requirements of the three significant drop orientations. The properties of regions I, II and III are selected to control the decelerations resulting from vertical, corner, and shallow drop orientations, respectively. The properties used to describe the stress-strain behavior of each of the three materials are discussed below. The dimensions A and B may vary for the limiters at each end of the packaging, but R0 and Ri are taken to be the same for both limiters. The same material properties are used for each of the limiters.

General Approach The ideal energy absorbing material is one that has a stress-strain curve that has a large strain region where the stress is constant. Such a material absorbs the maximum energy while minimizing force (which determines the magnitude of the deceleration).

Wood, foam, and honeycomb materials exhibit such behavior and are prime candidates for impact limiter crushable material. If the constant stress region of the stress-strain curve is of primary interest, the forces may be calculated as the crush stress times the area of the surface defined by the intersection of the target and the impact limiter. This approach assumes that the crush stress, which acts normal to the crush surface, is not influenced by stresses acting in directions parallel to the crush surface (i.e., the confining stresses). This assumption is made in the computer code. The crush stress used as input to the code is selected to represent that value which is consistent with the degree of confinement afforded by the impact limiter geometry for the drop orientation considered.

Therefore, the crushable material is modeled in the code with a one dimensional (oriented normal to the crush surface) stress-strain law. The properties of the stress-strain law are selected to represent the degree of confinement provided by stresses acting in the other two dimensions. The properties of the crushable material are not modified as the packaging rotates but are selected to represent the material properties for the initial crush direction of the material.

Page 2.10.8-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 A portion of the "crushed" area of the limiter is often not backed up by the packaging body (i.e., a projection of a point in this non backed up area normal to the target (impact surface) does not intersect the cask body). The user must specify the percentage of these forces which are to be included in the calculation. The confinement provided by the overall construction of the limiter will determine the extent to which these non backed up forces are actually effective. The computer code does not perform any computations which would allow the user to judge the adequacy of the selected percentage of non backed up forces which are counted.

The evaluation of the impact area and its centroid (required to locate the impact forces) is computationally complicated because of the many variations possible in the manner in which the target intersects the limiter. This problem is resolved by dividing the surface of the limiter into many small segments. The segment is located relative to the target at each computation. If the segment's original location is below the target, then it has crushed and it contributes a force equal to the stress times its area projected on the target. The location of this force is also known. The strain at the segment may also be evaluated so that the peak strains may be determined and stresses may be evaluated for strains which fall outside of the constant crush stress region of the stress-strain law.

The forces must be calculated at each time that the solution for the packaging response is computed. The problem, therefore, is to determine the forces acting on the limiters given the current location of the packaging center of gravity. The solution for the location of the packaging center of gravity is discussed in Section 2.10.8.8.1. The procedure used to perform these computations is as follows (each of the steps is detailed below).

1) Define the location of the target relative to the limiters from the current location of the packaging center of gravity relative to the target.

2) Divide the surface of the limiter into segments and calculate the strain in a one-dimensional element spanning the distance between the center of the segment and the packaging body.

3) Compute the stress in the element from the stress-strain relationship. Multiply the stress by the area of the element projected onto the target.

4) After all of the segments on the limiter are evaluated, sum the segment forces and moments of the forces to find the total force and moment acting on the packaging.

5) Calculate the horizontal force and moment of the horizontal force.

6) Use equations 2, 3, and 4 to extend the solution to the next time step. The new solution consists of the location of the packaging CG at the new time. The above steps are then repeated. This process is continued until the specified maximum time is reached.

Details of Force Computations Details of each of the six steps outlined above are given in this section. Note that the location of the packaging CG is known at the beginning of this computational sequence.

Page 2.10.8-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Deformation of the Limiter The first step in the computation is to evaluate the location of the limiters relative to the target given the location of the packaging CG relative to the target. The limiter position relative to the target is defined by the six variables, D1 through D6, as shown on Figure 2.10.8-9. The location of the cask at first contact is shown in Figure 2.10.8-9a with the subscript 0 added to the D's indicating initial values. The initial values of these parameters (when the lowest corner of the packaging first contacts) are found from the following geometric considerations.

D10 = 2R0 cos ,

D20 = 0 ,

D30 = B1 sin , (5)

D40 = D30 + D10 + L sin + B2 sin ,

D50 = D40 - D10 ,

D60 = D30 + L sin ,

At a given time t the packaging CG has displaced vertically u horizontally x and has rotated and reached the position shown in Figure 2.10.8-8b. Each of the six points have then fallen by an amount:

D = u + l [sin - sin( - )] + r [cos - cos( - )] (6)

Where l is the axial distance CG to point (+CG to top), and r is the radial distance CG to point (+CG to impact).

Then the corner deformation, D2, at time, t + l, becomes D2( t +1) = D2t + D2 .

Where l1 = l2 = -yL* - B1, l3 = -yL*

l4 = l5 = (l - y)L* + B2, l6 = (l - y)L*,

r1 = r4 = -R0, and r2 = r3 = r5 = r6 = R0.

Page 2.10.8-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 To facilitate the computation of strains in the limiter, the position of the limiter relative to the impact surface is classified as shown in Figure 2.10.8-10. There are three possible locations of the impact surface relative to the limiter. The task is therefore to define which of the three patterns apply, and to determine the parameters and in terms of the variables D1 through D6, just determined.

These deformations are next related to the three types of crush patterns for the bottom limiter shown on Figure 2.10.8-10. Crush pattern I applies when D1 < 0; D2 < 0; D3 > 0. (8)

Then, D2

= , and (9) cos D3 D2

= cos 1 B1 Crush pattern II applies when D1 > 0; D2 < 0; D3 > 0. (10)

Then, D2

= , and (11) cos D3 D2

= cos 1 .

B1 Crush pattern III applies when:

D1 > 0; D2 < 0; D3 < 0. (12)

Then, D2

= , and (13) sin D1 D2

= sin 1 .

2R0 The same set of equations applies to the top limiter if D1, D2, D3, and B1 are replaced with D4, D5, D6, and B2 in equations (8) through (13).

Page 2.10.8-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Strains in Limiters The next step in the computation is to calculate strains in the limiters given the deformation defined above. The limiters are first divided into segments as shown in Figure 2.10.8-11. The number of segments used for the bottom NB and the sides NS are input by the user. Locations on the surface of the limiters are described in terms of the (R, Z, ) coordinate systems shown on the figure. Strains in the segments along the sides of the limiters are calculated based on the location of the center of the segment (R0, Z, ). The segments at the bottom are divided into two pieces: one for R < Ri (i.e.

in Region 1) and the second for R > Ri. A strain is calculated for each of these two pieces for each segment along the bottom surface.

The strains are calculated as the deformation of the point normal to the crush surface divided by the undeformed distance of the point from the surface of the limiter to the outer container q, again measured normal to the crush surface. Therefore:

=/q (14)

Different equations govern each of these parameters for each of the three crush patterns as shown on Figure 2.10.8-10.

The geometry for crush pattern I is shown on Figure 2.10.8-12. Forces resulting from deformation of the side elements are neglected for this crush pattern. It may be shown that the deformation is:

= cos + (R cos - R0) sin (15)

The undeformed length of the element is taken measured to the plane of the packaging bottom so that q = A1 cos (16)

The geometry for crush pattern II is shown on Figure 2.10.8-13. The deformation of the points on the bottom (a) and along the side (b) may be represented with the same equation

= cos + (R cos - R0) sin - Z/cos (17)

The original length of the element depends on the intersection of the projection of the point on the impact surface with the outline of the limiter. These points are shown on Figure 2.10.8-13. The lengths are:

AZ q1 = ,

cos Page 2.10.8-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 X

q2 = , (18) sin BZ q3 = , and cos

(

q 4 = R02 R 2 sin 2 )1/ 2

+ R cos sin .

Where X = R cos + (R2 cos2 - R2 + R12)1/2.

The deformation for crush pattern III is shown on Figure 2.10.8-14. Deformations of points on the bottom of the limiter are neglected for this crush pattern. The deformation is Z R0 (1 cos )

tan

=

sin The original length is measured to Ri so that:

R0 R i q= . (20) sin Segment Stress The stresses in the elements are calculated from the above strains. As mentioned above, three sets of stress-strain laws are input to the code, one for each of the regions defined in Figure 2.10.8-8.

The location of the center of the segment on the surface of the limiter is used to determine which of the three stress-strain laws is to be used. The model may be viewed as a set of one dimensional rods which run from the center of the segment, normal to the target, to another boundary of the limiter. The entire rod is given the properties which the limiter material has at the beginning point of the rod (i.e., the intersection with the target).

The stress-strain law used for the materials is shown on Figure 2.10.8-15. Each of the seven parameters shown on the figure is input to the code for each of the three regions of the limiter. The arrows on the figure indicate the load-unload paths used in the model. The step in the crush strength is built into the stress-strain law so that two crushable materials in series may be modeled. The two crush strengths should be specified as the actual crush strengths of the two materials. The first locking strain L should be specified as the locking strain of the weaker material times the length of the weaker material divided by the total specimen length. The higher locking strain L should be specified as the first locking strain plus the locking strain of the stronger material times its length and divided by the specimen length.

Page 2.10.8-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 As stated above, the properties of the limiter material are not varied as the limiter crushes and the packaging rotates. Limiter materials such as wood exhibit anisotropic material properties. This must be accounted for when the properties are input to the code based on the anticipated direction of crushing. Most of the anisotropic wood data is based on tests performed in the elastic range. The following relationship has been used to represent wood properties for a loading which is applied at an angle with respect to the wood grain:

P1 cos 4 + P2 sin 4 P= (21) cos 4 + sin 4 Where P is the property of interest at angle , and P1 and P2 are properties parallel and perpendicular to grain.

Evaluation of Forces The stresses determined above are multiplied by the area of the segment projected onto the crush surface. The areas of the sidewall segments are (see Figure 2.10.8-11:

2R0 B cos( )

As = . (22)

(NB )(NS ) tan The area of the bottom segments is divided into two parts, one in region I and the other in region II. These areas are:

4R0 Lb sin( )

Ab = (23)

NB Where, Lb = (Ri2 - Rc2)1/2 for region 1, and Lb = (R02 - Rc2)1/2 - (Ri2 - Rc2)1/2 for region II.

These forces are summed for all of the elements to determine the total force acting on the packaging. The forces are also multiplied by their moment arms about the packaging CG to calculate the total moment acting on the packaging. The point on the segment is first projected, normal to the target, to evaluate whether or not it intersects the packaging body. If the projection does not intersect the packaging body, only a percentage of the force is included in the summation. The user specifies the percentage to be used.

Page 2.10.8-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Horizontal Force A horizontal force develops at the limiter/target interface. This force is only considered for the bottom limiter (i.e., the first to impact) since the packaging is always close to horizontal when the top impact limiter is in contact.

The horizontal force Fh is first calculated as that required to restrain horizontal motion of the tip of the limiter.

The horizontal acceleration at the tip of the bottom limiter (point 2 on Figure H

2.10.8-9) may be related to the CG motion of the packaging by:

H ( 1 )

= x L* + B cos + R sin 0 (24)

Where = + .

2 Equating H to zero would result in no acceleration of the tip in the horizontal direction and provides the solution for x in terms of .

Substituting this solution for x into Equation (3) results in an expression for the horizontal force Fh required to restrict horizontal acceleration of the tip, in terms of the rotational acceleration . Finally, equation 4 is used to eliminate with the following result.

Fh =

( )

MvW L* + B1 cos + R0 sin

+ W (L + B )cos + R sin (25)

2 Jg 1 0 Where Mv is the moment due to vertical forces, which is equal to Fv1Xv1 - Fv2Xv2, and W is the packaging weight.

This force is restricted to:

Fh < Fv1 (26)

Where is the coefficient of friction specified by user.

Page 2.10.8-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.8.9 References

1. Federal Specification MMM-A-188b.

2. Dreisback, J.F., Balsa Wood and Its Properties, Columbia Graphs, Columbia, CT 1952.

3. Marks Standard Handbook for Mechanical Engineers, Eighth Edition, pg. 6-124.

4. American Society Of Mechanical Engineers, ASME Boiler And Pressure Vessel Code,Section III, Appendices, 1989.

5. 10 CFR 71, Packaging And Transportation Of Radioactive Material.

6. Transnuclear ADOC Computer Program, Rev. 1.

Page 2.10.8-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-1 Mechanical Properties of Wood and Wood Adhesive Minimum Properties of Adhesive [1]

Shear Strength by Compression Loading 2,800 lb in2 Shear Strength by Tension Loading 340 lb in2 Properties of Heavy Balsa (10-12 lb ft3) [2]

Shear Strength Parallel to Grain 315-385 psi max.

Tensile Strength Perpendicular to Grain 140-160 psi Properties of Redwood [3]

Shear Strength Parallel to Grain 940 psi Tensile Strength Perpendicular to Grain 240 psi Page 2.10.8-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-2 Typical Wood Material Properties Property High Density Balsa Redwood Density 10-12 lb ft3 18.7-27.5 lb ft3 Parallel to Grain Crush Stress 1560-2010 psi 5000-6500 psi Locking Strain 0.8 0.6 Unloading Modulus 32,000 psi 1,247,000 psi Locking Modulus 10 (max. crush stress) 10 (max. crush stress)

Perpendicular to Grain Crush Stress 300-420 psi 750-975 psi Locking Strain 0.8 0.6 Unloading Modulus 32,000 psi 1,247,000 psi Locking Modulus 10 (max. crush stress) 10 (max. crush stress)

Page 2.10.8-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-3 First Impact Maximum Inertia g-Load versus Initial Angle of Impact for 30 Foot Drop, Using Maximum Wood Crush Stress Properties Impact Maximum g-Load During First Impact (Bottom),

Angle, Maximum Wood Properties 30 Foot Axial Transverse Drop CG Top Bottom CG 0 4 51 50 51 5 7 gnor = 27

gnor = 24 27 grot = 52* grot = 35 10 10 gnor = 27* gnor = 27 27 grot = 51* grot = 39 15 15 gnor = 28* gnor = 32 32 grot = 52* grot = 44 20 22 gnor = 27* gnor = 39 39 grot = 49* grot = 51 30 26 gnor = 33* gnor = 32 33 grot = 57* grot = 35 40 21 3 39 18 45 27 3 36 19 50 29 8 25 17 60 29 8 15 11 63.8 32 8 14 11 70 38 8 12 10 80 44 4 7 5 90 49 3 1 2

Maximum acceleration occurred during second impact.

Page 2.10.8-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-4 First Impact Maximum Inertia g-Load versus Initial Angle of Impact for 30 Foot Drop, Using Minimum Wood Crush Stress Properties Impact Maximum g-Load During First Impact (Bottom),

Angle, Minimum Wood Properties 30 Foot Axial Transverse Drop CG Top Bottom CG 0 3 39 38 39 5 5 gnor = 25

gnor = 21 28 grot = 47* grot = 29 10 8 gnor = 23* gnor = 22 23 grot = 45* grot = 35 15 12 gnor = 22* gnor = 26 26 grot = 42* grot = 36 20 19 gnor = 23* gnor = 32 32 grot = 43* grot = 44 30 21 gnor = 27* gnor = 26 27 grot = 48* grot = 28 40 20 4 37 20 45 24 3 32 17 50 25 6 22 14 60 26 7 13 10 63.8 29 7 13 10 70 38 7 12 10 80 46 6 8 7 90 37 4 1 2

Maximum acceleration occurred during second impact Page 2.10.8-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-5 Maximum Impact Limiter Deformation versus Initial Angle of Impact, for 30 Foot Drop, using the Maximum Wood Properties Impact Limiter Deformation using Maximum Wood Properties Impact First Impact (bottom) Second Impact (top)

Angle Maximum Wood Maximum Maximum Wood Maximum 30 ft Drop Crush Depth (in.) Wood Crush Depth (in.) Wood Strain Strain 0 13.58 0.462 13.42 0.451 5 13.17 0.442 16.17 0.562 10 15.75 0.529 15.52 0.537 15 18.72 0.617 15.16 0.521 20 22.26 < 0.389* 15.50 0.531 30 29.77 < 0.463

19.00 < 0.311*

40 41.17 0.387 - -

45 37.09 0.440 - -

50 35.18 0.541 - -

60 33.00 0.593 - -

63.8 30.88 0.566 - -

70 30.63 0.615 - -

80 21.46 0.468 - -

90 10.23 0.275 - -

Value computed using more accurate method than ADOC. All other strain values not noted are taken from the ADOC results files which are very conservative.

Page 2.10.8-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-6 Maximum Impact Limiter Deformation versus Initial Angle of Impact, for 30 Foot Drop, using the Minimum Wood Properties Impact Limiter Deformation using Minimum Wood Properties First Impact (bottom) Second Impact (top)

Impact Angle Maximum Maximum Maximum Wood Maximum Wood 30 ft Drop Wood Wood Crush Depth (in.) Crush Depth (in.)

Strain Strain 0 15.87 0.552 15.72 0.546 5 14.72 0.503 18.89 0.669***

10 17.19 0.586 18.15 0.640 15 20.06 0.400* 17.80 0.625 20 23.48 0.389* 18.07 0.574*

30 31.78 0.463* 20.88 0.311*

40 33.66 0.837 - -

45 41.25 0.486 - -

50 39.30 0.603 - -

60 37.13 0.672** - -

63.8 34.60 0.644 - -

70 33.47 0.681 - -

80 23.95 0.532 - -

90 13.16 0.353 - -

Value computed using more accurate method than ADOC. All other strain values not noted are taken from the ADOC results files which are very conservative.

- The maximum strain is slightly higher (0.672 - 0.630 = 0.042) than the maximum locking strain of 0.630 (in Region II). However, the region of maximum wood strain is very small and will therefore, not affect the impact acceleration or wood mechanical properties significantly.

- - The maximum strain is slightly higher than the maximum locking strain of 0.651 (in Region III).

However, the region of maximum wood strain is very small and will therefore, not affect the impact acceleration or wood mechanical properties significantly.

Page 2.10.8-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-7 Maximum Inertial g-Load during 1 Foot Drop Impact Maximum g-Load During First Impact, Angle, Maximum Wood Properties 1 foot Axial Transverse Drop CG Top Bottom CG 90 12 0 0 0 0 1 16 16 16 63.8 5 2 3 3 Impact Maximum g-Load During First Impact, Angle, Minimum Wood Properties 1 foot Axial Transverse Drop CG Top Bottom CG 90 9 1 0 0 0 1 12 12 12 63.8 5 2 3 2 Page 2.10.8-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-8 Maximum Impact Limiter Deformation versus Initial Angle of Impact for 1 Foot Drop Maximum Wood Properties First Impact Second Impact Impact Angle Maximum Wood Maximum Maximum Maximum Wood 1 ft Drop Crush Depth Wood Wood Crush Depth (in.)

(in.) Strain Strain 90° 1.17 0.032 - -

0° 2.77 0.038 2.75 0.038 63.8° 10.71 0.177 - -

Minimum Wood Properties First Impact Second Impact Impact Angle Maximum Wood Maximum Maximum Wood Maximum 1 ft Drop Crush Depth Wood Crush Depth Wood (in.) Strain (in.) Strain 90° 1.71 0.046 - -

0° 3.06 0.050 3.05 0.049 63.8° 12.06 0.202 - -

Page 2.10.8-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-9 Loading Used in Cask Body Analysis, Appendix 2.10.1 versus Maximum g-Loads Predicted by ADOC Program Accident Conditions (30 Foot Drops)

Drop Orientation Max. g-Load from ADOC Input Loading Used in FEA End Drop 49g Axial 49g Axial on Lid and Bottom Side Drop 51g Transverse 51g Transverse CG over Corner Drop 32g Axial 32g Axial on Lid And Bottom (63.8°) 14g Transverse 14g Transverse 20° Slap Down 22g Axial 22g Axial on Top Impact Limiter 39g Transverse (normal) 39g Transverse (normal)

Normal Conditions (1 Foot Drops)

Drop Orientation Max. g-Load from ADOC Input Loading Used in FEA 90 End Drop 12g Axial 12g Axial on Lid and Bottom 0 Side Drop 16g Transverse 16g Transverse Page 2.10.8-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.8-10 Loading Used in Basket Structural Analysis, Appendix 2.10.5 versus Maximum g-Load Predicted by ADOC Program Accident Conditions (30 Foot Drops)

Maximum g-Load, Input g-Load Used in Basket Structural Drop Orientation from ADOC Analysis, Including Dynamic Load Factor 75 g Axial 90 End Drop 49 g Axial (Conservatively Using Higher g-load) 75 g Transverse 0 Side Drop 51 g Transverse (Conservatively Using Higher g-load)

Normal Conditions (1 Foot Drops)

Maximum g-Load Input g-Load Used in Basket Structural Drop Orientation from ADOC Analysis, Including Dynamic Load Factor 90 End Drop 12 g Axial 20 g Axial 0 Side Drop 16 g Transverse 20 g Transverse Page 2.10.8-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-1 Impact Limiter Geometry Page 2.10.8-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-2 Sample Force/Deflection Curve for Balsa Page 2.10.8-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-3 Sample Force/Deflection Curve for Redwood Page 2.10.8-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 B regions are bolsa.

R regions are redwood.

Dashed lines indicate grain orientation.

Figure 2.10.8-4 ADOC Computer Model for TN-40 Transport Package Page 2.10.8-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-5 TN-40 Package Geometry during Impact for Wood Strain Computation Page 2.10.8-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-6 Geometry of Packaging Page 2.10.8-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-7 Packaging at Time, t Page 2.10.8-49

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-8 Geometry of Impact Limiter Parameters Page 2.10.8-50

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-9 Definition of Limiter Deformation Page 2.10.8-51

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-10 Crush Pattern in Impact Limiter Page 2.10.8-52

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-11 Impact Limiter Segments Page 2.10.8-53

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-12 Strain Computation for Crush Pattern I Page 2.10.8-54

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-13 Strain Computation for Crush Pattern II Page 2.10.8-55

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-14 Strain Computation for Crush Pattern III Page 2.10.8-56

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-15 Wood Stress-Strain Curve Page 2.10.8-57

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Where:

R = 12,500 kips

µ = 0.42 x1 = 12.00/2 = 6.00 in.

x2 = (I.L. o.d. - Cask o.d.)/2 - crush depth

= (144.00 - 91.00)/2 - 5.62 = 20.88 in.

x3 = 12.00 in. - 11.16 in. = 0.84 in.

Figure 2.10.8-16 Impact Limiter Free Body Diagram during 20 Slap Down Page 2.10.8-58

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-17 Impact limiter Lifting Lug Geometry Page 2.10.8-59

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.8-18 Cask Geometry during Tip-Over Event Page 2.10.8-60

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.9 TABLE OF CONTENTS 2.10.9.1 Introduction........................................................................ 2.10.9-1 2.10.9.2 Scaling Relationships ........................................................ 2.10.9-1 2.10.9.3 Test Article Description...................................................... 2.10.9-2 2.10.9.4 Test Description ................................................................ 2.10.9-3 2.10.9.5 Data Measurement ............................................................ 2.10.9-7 2.10.9.6 Test Data and Results ....................................................... 2.10.9-8 2.10.9.7 Conclusions ..................................................................... 2.10.9-16 2.10.9.8 References ...................................................................... 2.10.9-17 LIST OF TABLES Table 2.10.9-1 Summary of the Test Results................................................ 2.10.9-18 Table 2.10.9-2 (deleted)................................................................................ 2.10.9-19 LIST OF FIGURES Figure 2.10.9-1 One-Third Scale Test Article ................................................. 2.10.9-20 Figure 2.10.9-2 Accelerometer Locations ...................................................... 2.10.9-21 Figure 2.10.9-3 0° Side Drop Test Setup ....................................................... 2.10.9-22 Figure 2.10.9-4 CG Over Corner Drop Test Setup ......................................... 2.10.9-23 Figure 2.10.9-5 20° Slap Down Test Setup .................................................... 2.10.9-24 Figure 2.10.9-6 90° End Drop Test Setup ...................................................... 2.10.9-25 Figure 2.10.9-7 Puncture Drop Test Setup .................................................... 2.10.9-26 Figure 2.10.9-8 Test Article and Accelerometer Locations............................. 2.10.9-27 Figure 2.10.9-9 0° Side Drop Test Rigging .................................................... 2.10.9-28 Figure 2.10.9-10 Acceleration Time History, 0° Side Drop ............................... 2.10.9-29 Figure 2.10.9-11 Test Article After 0° Side Drop .............................................. 2.10.9-30 Figure 2.10.9-12 Test Article After 0° Side Drop, Upper End Details ............... 2.10.9-31 Page 2.10.9-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-13 Test Article After 0° Side Drop, Lower End Details ............... 2.10.9-32 Figure 2.10.9-14 TN-40 Impact Limiter Shell Damage After 0° Side Drop ....... 2.10.9-33 Figure 2.10.9-15 CG Over Corner Drop Test Rigging ...................................... 2.10.9-34 Figure 2.10.9-16 CG Over Corner Drop Acceleration Time History ................. 2.10.9-35 Figure 2.10.9-17 CG Over Corner Drop Impacted Surface Damage ............... 2.10.9-36 Figure 2.10.9-18 CG Over Corner Drop Impact Limiter Deformation ............... 2.10.9-37 Figure 2.10.9-19 20° Slap Down Test Rigging ................................................. 2.10.9-38 Figure 2.10.9-20 20° Slap Down Drop Acceleration Time History.................... 2.10.9-39 Figure 2.10.9-21 Test Article After 20° Slap Down Drop .................................. 2.10.9-40 Figure 2.10.9-22 Impact Limiter (Slap Down End) After 20° Slap Down Drop . 2.10.9-41 Figure 2.10.9-23 90° End Drop Test Rigging ................................................... 2.10.9-42 Figure 2.10.9-24 Acceleration Time History, 90° End Drop ............................. 2.10.9-43 Figure 2.10.9-25 Test Dummy and Chilled Impact Limiter After Initial 90° End Drop............................................................................... 2.10.9-44 Figure 2.10.9-26 Test Dummy and Chilled Impact Limiter Post-Drop Detail .... 2.10.9-45 Figure 2.10.9-27 Puncture Drop Test Rigging.................................................. 2.10.9-46 Figure 2.10.9-28 Test Article After Puncture Drop ........................................... 2.10.9-47 Figure 2.10.9-29 Puncture Pin Damage to Impact Limiter ............................... 2.10.9-48 Page 2.10.9-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 TN-40 PACKAGE IMPACT LIMITER TESTING 2.10.9.1 Introduction A series of dynamic tests have been performed on one-third scale models of the TN-40 transport package impact limiters. The tests were performed to evaluate the effects of the 30 foot free drop hypothetical accident defined in 10 CFR 71.73(c)(1) [1]. The test results are used to verify the analyses performed for the TN-40 transport package. The objectives of the TN-40 impact limiter test program are:

Demonstrate that the inertia g values and forces calculated in Appendix 2.10.8 and used in the analyses presented in Appendices 2.10.1 through 2.10.7 are adequate,

Demonstrate that the extent of crush depths is acceptable, i.e., limiters do not bottom out and the neutron shield or trunnions do not impact the target,

Demonstrate the adequacy of the impact limiter enclosure,

Demonstrate adequacy of the impact limiter attachment design,

Evaluate the effects of low temperature (-20 °F) on the crush strength and dynamic performance of the impact limiters, and,

Evaluate the effects (puncture depth and shell damage) of a 40 inch drop onto a scaled six inch diameter puncture bar on a previously crushed impact limiter, as per 10 CFR 71.73(c)(3).

2.10.9.2 Scaling Relationships The scale models of the TN-40 cask and impact limiter are constructed with a geometric scale factor of 1/ = 1/3. As shown in Appendix B of Reference [2], the following scale factors apply.

Length:

Lp = Lm Surface area:

Ap = 2 Am Moment of inertia:

Ip = 4 Im Section modulus:

Sp = 3 Sm Page 2.10.9-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Weight:

Wp = 3 Wm Energy absorbed during drop (from same height h):

Ep = Wp h = 3 Wm h = 3 Em Velocity at beginning of impact:

Vp = (2gh) 1/ 2

= Vm Where; is the scale factor, the subscript p refers to the full size, and the subscript m refers to the model.

During impact, the impact limiter materials will deform or crush. Since the model and full size impact limiters are made of the same materials, they deform under the same

stress, Sp = Sm.

Therefore we have the following relationships:

Force during impact:

Fp = Sp Ap = Sm 2 Am = 2 Fm Deformation:

Dp = Ep /Fp = 3 Em /2 Fm = Dm Impact duration:

Tp = Dp /Vp = Dm /Vm = Tm Impact deceleration:

ap = Vp /Tp = Vm / Tm = 1/ am 2.10.9.3 Test Article Description The test article for the dynamic tests consists of a solid carbon steel test body with an impact limiter on each end. The test article, shown in Figure 2.10.9-1, is constructed to be as close as possible to one-third of the full size packaging.

Page 2.10.9-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The impact limiters are attached to each other by thirteen 0.5 inch diameter tie rods, tightened snug tight. Each limiter is also fastened to the test dummy with four 0.5 inch bolts.

The test article weighs approximately 10,100 lb (the total weight of the full-size package is 271,460 lb), and has maximum dimensions of approximately 87.0 inches long by 48.0 inches in diameter (The full size dimensions are 260.87 inches long and 144.0 inches in diameter).

The test body and each impact limiter are equipped with lifting lugs to facilitate lifting during handling and testing.

2.10.9.4 Test Description 2.10.9.4.1 Equipment and Instrumentation The drop testing was performed at the National Technical Systems (NTS) facility located at Acton, Massachusetts. The drop testing was performed in accordance with approved written procedures.

Lifting and dropping the test article was accomplished using a mobile crane. A quick release mechanism was used to initiate the drop. It consisted of a hydraulic piston that loaded a bolt to failure releasing a shackle supporting the test article via a rigging system. The rigging system consisted of nylon straps and padded shackles in order to minimize damage to the accelerometers installed on the test dummy.

An inclinometer was used to measure the initial angle ( 1°) of the test body longitudinal axis with respect to the drop pad (i.e., impact surface). A measured line, 30 feet long

(+ 3.0, -0.0 inches), was attached to the lowest point on the test package in order to assure the proper drop height.

The impact surface was a 2 in. thick steel plate attached to a concrete block weighing approximately 250,000 lb. resting on bedrock. This configuration can be considered as an essentially unyielding surface.

A puncture bar made of cold-rolled steel was welded to the impact surface for the 40 in.

puncture drop. The pin was scaled to match the test article resulting in a 2 in. diameter pin with the upper end edges rounded to a radius of approximately 0.083 in.

Accelerometers were used to measure the impact g load for all drops performed.

Twelve PCB Piezotronics 353B18 accelerometers were attached to aluminum blocks which were bolted to the test body at 0°, 90°, 180°, and 270° orientations at three elevations: the approximate center of gravity location and adjacent to each impact limiter. The 12 accelerometer locations are shown in Figure 2.10.9-2.

Page 2.10.9-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 PCB Model 353B18 quartz shear accelerometers were used to measure the test dummy response. These transducers have a measurement range of +/- 500 g, with a nominal frequency range of 1 - 10,000 Hz (+/- 5%). The accelerometers were connected to a Spectral Dynamics, Inc. Puma System signal analyzer. The accelerometer responses were recorded digitally and processed after completion of the test.

The output signals were filtered using a 600Hz low pass filter to remove the higher frequencies present in the data. The high frequencies represent vibrations of the test dummy due to small displacements (low stresses), which excite the accelerometers and tend to mask the low frequency rigid body acceleration. This low frequency acceleration is masked, because both low frequency rigid body and high frequency natural vibration accelerations are superimposed and the net acceleration is recorded. Filtering the data is necessary to remove these high frequency accelerations.

2.10.9.4.2 Drop Test Orientations A. Pre Test Orientations The four 1/3 scale impact limiters that were used for the test are identified as 1, 2, 3, and 4. The drop test orientations were performed in the following sequence.

Drop Test Original Sequence Impact Test Drop Drop Limiter Impact Number Orientation Height Number Sequence Comments 0° 1 - Limiters 1 and 2 were installed 1 30 feet Side Drop 2 -

1 1st The test article was rotated 64° 180° so the undamaged 2 CG Over 30 feet 2 2nd portion of the impact limiters Corner Drop contacts the impact surface.

3 1st Limiters 1 and 2 were removed 20° 3 30 feet and replaced with limiters 3 Slap Down 4 2nd and 4 before the drop.

3 1st Limiter 3 was removed and 90° chilled at -20 °F for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4 30 feet End Drop 4 None before reinstallation on the test dummy.

90° 3 1st Drop onto 2 inch diameter End Drop puncture bar.

5 40 inches (Puncture 4 None Test)

The 0° side drop was performed to obtain the highest transverse acceleration and to demonstrate that no portion of the test dummy (including trunnions) impacted the surface of the pad.

Page 2.10.9-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The 64° CG over corner drop was performed to predict the maximum crush distance from the corner of the impact limiter and to demonstrate the corners of the test dummy were protected by the impact limiter.

The 20° slap down drop was performed to demonstrate that the impact limiters stay attached to the test dummy because the second impact (slap down) puts the highest load on the impact limiter attachments, and impact limiter stainless steel shell.

The 90° end drop orientation was performed to obtain the highest axial deceleration. For the 90° end drop, the bottom impact limiter (3) was chilled at -20 °F for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in order to acquire the most conservative estimate of the axial g load.

A 40 inch drop onto a 1/3 scale 6 inch diameter puncture bar was performed in accordance with 10 CFR 71.73(c)(3) in order to evaluate the effects of this drop on the TN-40 transport package. The test article was dropped in the 90 end drop orientation onto the puncture bar subsequent to the 30 foot end drop. The test article was positioned so that it impacted the puncture bar close to its axial centerline and thus near the center of gravity. This orientation was chosen because it assures that the puncture impact absorbs nearly 100% of the drop energy. Also the center of the impact limiter outer plate, where the puncture impact occurs, is the weakest portion of the impact limiter since there are no gussets in this location.

B. Modified Drop Tests After the initial drop test (0° side drop), the accelerometer data was determined to be unusable because of an auto-scaling feature of the data acquisition system that did not perform as expected. After NTS reset the system for manual scaling the second planned drop was conducted. This was the 64° CG over corner drop using the same pair of impact limiters. After the CG over corner drop test a second 0° side drop test was performed. The same pair of impact limiters was used. The test package was rotated 90° such that the undamaged portion of the impact limiter would contact the impact surface, thus providing undamaged crush material. The accelerometers were left in their original locations, but since they were equally spaced at four locations around the test body, the desired impact accelerations could still be obtained after reorienting the accelerometers to be perpendicular with the impact surface.

After the 90° end drop examination of the test data revealed results that were inexplicable. Several data channels showed accelerations that were opposite sign from the expected values and had significant differences in magnitude. In order to salvage useful data from the existing test hardware, a second end drop was performed at a later date using the least damaged test limiter for the impacted end drop limiter. The limiter was not chilled, however, since the environmental control of the wood had been compromised due to previous test damage.

The original sequence of the tests shown in A above was modified to list all the drop tests that were performed:

Page 2.10.9-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Modified Drop Test Sequence Impact Test Drop Drop Limiter Impact Number Orientation Height Number Sequence Comments 0° 1 - Limiters 1 and 2 installed.

1 30 feet Side Drop 2 -

1 1st The 1 and 2 impact limiters 64° were rotated 180° so the 2 CG Over 30 feet 2 2nd undamaged portion of the Corner drop impact limiters face the pad.

1 - The test body and limiters 0° were rotated 90° so that an 3 Side Drop 30 feet 2 - undamaged portion of the (2nd test) impact limiters faced the pad.

3 1st Limiters 1 and 2 were 20° 4 30 feet removed and replaced with Slap Down 4 2nd limiters 3 and 4.

3 1st Limiter 3 was removed and 90° chilled at -20 °F for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 5 30 feet End Drop 4 - before being re-installed on the test body.

90° 3 1st Drop onto 2 inch diameter End Drop puncture bar.

6 40 inches (Puncture 4 -

Test) 2 1st Limiters 2 and 4 were used.

90° The center portion of limiter 2 was relatively undamaged by 7 End Drop 30 feet 4 - previous drops and thus (2nd test) provided a useable crush volume.

The test setup for the 0° side drop is shown in Figure 2.10.9-3. For the side drop test, the accelerometers were oriented to measure accelerations in the drop direction (perpendicular to the drop pad surface).

The test setup for the 64° CG over corner drop is shown in Figure 2.10.9-4. The accelerometers located along the center of gravity and near the top and bottom impact limiter (1st impact) were oriented to measure accelerations 64° from the axis of the test model (perpendicular to the drop pad surface when the test model is oriented at a 64° angle). The other accelerometers were oriented to measure accelerations parallel to the test model axis.

The same pair of impact limiters was rotated 90° so the undamaged portion of the limiter would be face down for the second side drop (test number 3). The accelerometers were re-oriented to measure accelerations in the drop direction (perpendicular to the impact surface). The test setup is the same as shown in Figure 2.10.9-3 except for the 90° rotation of the test article.

Page 2.10.9-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The test setup for the 20° slap down drop is shown in Figure 2.10.9-5. The accelerometers located along the side of the test dummy were oriented to measure accelerations 70 from the axis of the test model (perpendicular to the drop pad surface when the test model is oriented at a 20° angle). The accelerometers along the top and bottom of the test dummy were oriented to measure accelerations perpendicular to the test model axis (perpendicular to the drop pad surface during slap down when the test modal axis is parallel to the impact surface).

The test setup for the 90° end drop is shown in Figure 2.10.9-6. The package was oriented with the test dummy bottom facing down so that the impact occurred on the bottom end of the package. For the end drop test, the accelerometers were oriented to measure accelerations in the drop (axial) direction. The bottom impact limiter (3) was kept in a temperature chamber held at a temperature of -20 °F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months immediately prior to the test. The time between removal of the impact limiter from the conditioning chamber and the drop was approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Examination of the test data from the first 90° end drop revealed results that were inexplicable. Several data channels showed accelerations that were opposite sign from the expected values and had significant differences in magnitude. In order to extract useful data from the existing test hardware, a second 90° end drop test (test number 7) was performed at a later date using the least-damaged test impact limiter for the impacted end drop. The limiter was not chilled -20 °F, however, since the environmental control of the wood had been compromised due to previous test damage.

Impact limiters 2 and 4 were reused for the second 90° end drop test (test number 7).

The location and orientations of the accelerometers were same as for the first 90° end drop test. The purpose of this test was to predict the g load at the room temperature.

The test setup was the same as for the first end drop and is shown in Figure 2.10.9-6.

The test setup for the 90° puncture drop is shown in Figure 2.10.9-7. During the puncture drop the package was oriented so that the puncture bar impacted on the bottom end of the package. A 2 inch diameter solid cylindrical puncture bar, 18 inches long was used. The puncture bar was constructed from mild steel and was welded to the drop pad with its long axis oriented in the vertical direction. Accelerometer data was not taken during the puncture drop test.

A photograph of the accelerometer locations for each channel is shown in Figure 2.10.9-8. Accelerometers 1-4 are on the front (left end of test dummy), 5-8 are in the middle, and 9-12 are on the right end of the test dummy. Note that accelerometers 4, 8 and 12 are not visible in Figure 2.10.9-8.

2.10.9.5 Data Measurement The following data was measured and recorded before, during, and after each drop test.

1) Prior to each drop test

a. Torque of the impact limiter bolts.

Page 2.10.9-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

b. Impact limiter dimensions.

c. Height from test article to drop pad.

d. Angular orientation of the test article to the impact surface.

e. Atmospheric condition data, i.e., ambient temperature, wind speed, immediately and prior to the release of the test article.

2) During each drop test

a. Test article behavior on videotape.

b. Date and time of test.

c. Observations of damage or unexpected behavior of the test article

d. Impact acceleration time histories (excluding the puncture drop test).

3) Following each drop test

a. Observations of the damage to the test article on features other than the limiters, i.e., attachment bolts.

b. Measurements of deformation to each impact limiter to fully describe the extent of the damage. These measurements include:

i. Depth of internal and external crush of the impact limiter.

ii. Overall thickness of each impact limiter after each test.

iii. Dimensions of impact footprint.

2.10.9.6 Test Data and Results For purposes of reviewing test results, it should be noted that the energy to be absorbed by the scale model is approximately 1/27 of the full scale TN-40 package energy. The acceleration of the model is approximately three times that of the full size cask, and the crush deformation of the model limiter is approximately one-third that of the full size limiter. The impact force applied to the model is determined by multiplying the mass by the rigid body acceleration (F = ma). The model force is 1/9 of the full scale force.

2.10.9.6.1 0° Side Drop Test The first drop test performed was the 0° side drop. Impact limiters 1 and 2 were installed on the test dummy. Two lifting straps were used to connect the test article to a padded shackle that attached the test article to the release mechanism. Figure 2.10.9-9 is a photograph of the test configuration prior to the 0° drop.

Page 2.10.9-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Accelerometer Data As described in the previous section that the acceleration time history data for the first 0° side drop test was lost due to a problem with the self-scaling routine of the data acquisition software. After resolving the scaling problems, a second 0° side drop was performed. The acceleration time histories from channels 2, 6, and 10 were selected as providing the highest accelerations along the length of the test dummy. The plots generally show a single rounded peak roughly 0.02 sec. long, with a high frequency low amplitude signal superimposed on top of it.

The following table shows the maximum transverse accelerations measured by the accelerometers during the second 0° side drop (converted to full scale), as well as the maximum acceleration predicted by ADOC as described in Appendix 2.10.8.

Predicted Maximum Measured Transverse Average Measured Transverse Acceleration Accelerometer Acceleration (gs) Transverse (gs)

Location (converted to full scale) Acceleration (gs) (Appendix 2.10.8)

Top (2) 68 Center of Gravity (6) 50 57 51 Bottom (10) 55 The accelerations measured during the side drop are higher than predicted by the ADOC computer program. The acceleration results presented in the above table are taken from the measured acceleration data filtered with a 600 Hz low pass filter. Figure 2.10.9-10 shows the filtered acceleration time history from accelerometer 6, which is characteristic of the acceleration plots in general. Note that the acceleration plotted in Figure 2.10.9-10 is for the 1/3 scale package, and thus is equivalent to 3 times the full scale accelerations.

Crush Depth Measurements After the first 0° side drop test, crush depths of the impact limiters were measured. Even though acceleration data from the first test was not useable, the crush values are valid.

There was evidence of both inside and outside crushing. The following table summarizes the measured and predicted crush depths for the bottom impact limiter. A spring back of 0.50 inches is assumed (based on previous crush tests). See Appendix 2.10.4 of Reference [3].

Impact Limiter Impact Limiter Number 1 Number 2 Maximum Inside Crush Depth (in.) 1.44 1.50 Maximum Outside Crush Depth (in.) 0.75 0.75 Spring Back (in.) 0.50 0.50 Total Crush Depth (in.) 2.69 2.75 Predicted Crush Depth 1/3 (in.) (From ADOC) 4.52 Page 2.10.9-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 From the above table it can be seen that the measured crush depths are slightly less that those predicted by the ADOC computer program. Note that the actual scale test measurements are used for all deflection evaluations, with the ADOC results scaled to the 1/3 scale test.

It should also be noted that neither the neutron shield nor the trunnions would contact the impact surface during the impact. The distance between the outer diameter of the neutron shield and the outside diameter of the impact limiter is 7.16 in. Therefore, a clearance of 7.16 - 2.75 = 4.41 in. would remain between the impact surface and the neutron shield, based on the measured crush depth. Similarly, a distance of 3.84 in.

would remain between a trunnion and the impact surface.

Damage Assessment Both impact limiters remained attached to the test dummy during and after the side drop impact. All of the tie rods and tie rod brackets remained intact, thus preventing separation of the impact limiters from the test dummy. In addition, the impact limiter attachment bolts remained in place, in spite of damage to two of the eight bolting brackets.

Only a single small opening in the stainless steel shell of each of the impact limiters was evident. Both openings consisted of a tear along the weld between two of the outer flat plates of the impact limiter. The tears were roughly 4 inches long. Despite these tears, all impact limiter wood remained completely confined within the shell.

Figure 2.10.9-11 through Figure 2.10.9-14 are photographs of the test article after the 0° side drop.

2.10.9.6.2 64° CG over Corner Test The second drop test performed was the 64° CG over corner test. Impact limiters 1 and 2 were again used. The test article was rotated about its longitudinal axis 180° so that an undamaged portion of the limiter was exposed to the impact surface. The rigging for this test is shown in Figure 2.10.9-15.

Accelerometer Data Accelerometer results from accelerometers 9, 10, 11 and 12 at the upper end of the test dummy were used to evaluate this drop orientation. These represented the four highest accelerations. The output from accelerometer 10 is shown in Figure 2.10.9-16.

Accelerometers 9 and 11 were oriented parallel with the test dummy axis while accelerometers 10 and 12 were adjusted so that their axes were parallel with the drop (i.e., perpendicular to the test pad surface). The average acceleration of the 4 accelerometers in the direction of the test dummy axis is 34 gs. This includes correcting the values of accelerometers 10 and 12 to account for their orientation. The predicted acceleration is taken from ADOC results given in Appendix 2.10.8. Note that the predicted acceleration value is slightly less than the measured value.

Page 2.10.9-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Measured Versus Predicted Accelerations Measured Acceleration Average Predicted (gs) Measured Maximum Accelerometer Location (converted to full Acceleration Acceleration (see Figure 2.10.9-2) scale) (gs) (gs) 9 28 Axial Acceleration 10 37 34 32 (1st Impact) 11 39 12 33 Crush Depth Measurements The crush depths of the impact limiters were measured after the CG over corner drop.

The following table summarizes the measured and predicted crush depths for the bottom impact limiter. A spring back of 0.50 inches is assumed (based on previous crush tests). See Appendix 2.10.4 of Reference [3].

Impact Limiter 1

Maximum Inside Crush Depth (in.) 0.0 Maximum Outside Crush Depth (in.) 8.0 Spring Back 0.5 Total Crush Depth (in.) 8.5 Predicted Crush Depth 1/3 (in.)

(From ADOC) 10.3 From the above table it can be seen that the measured crush depths are slightly less that those predicted by the ADOC computer program.

Damage Assessment The primary purpose of the CG over corner drop test is to demonstrate that the impact limiter has sufficient material to protect the corner of the test dummy in this orientation and to demonstrate the adequacy of the impact limiter attachment design.

Both impact limiters remained attached to the test dummy during and after the CG over corner drop impact. All of the tie rods, and all but one of the tie rod brackets remained intact, thus preventing separation of the impact limiters from the test dummy. In addition, the impact limiter attachment bolts and brackets, although damaged, prevented the test dummy from separating from either impact limiter.

Several tears along welds on the impacted surface were noted. However, all of the shell plates remained in place and therefore would prevent significant loss on wood during a post-drop fire accident. The tears varied between 2 and 8 inches long. Despite these tears, all impact limiter wood remained completely confined within the shell.

Page 2.10.9-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-17 and Figure 2.10.9-18 are photographs of the test article after the CG over corner drop.

2.10.9.6.3 20° Slap Down Test The 20° slap down drop test was performed using newly installed impact limiters 3 and

4. The test article was oriented as shown in Figure 2.10.9-5. A two point strap rigging system was used to lift the test model by two lifting lugs. The two legs of the rigging system join at a single point that was shackled to the quick release mechanism. Figure 2.10.9-19 is a photograph of the test package rigging just prior to the 20° slap down drop.

Accelerometer Data The slap down event consists of two distinct impacts. The initial impact is the smaller of the two as it only stops the leading end of the test article and converts much of its kinetic energy from linear to rotational. The second impact is more severe in that the velocity of the second impact limiter is greater than that resulting from a 30 foot drop due to the added rotational velocity. The impact limiter attachment design is also exposed to unique loads due to the centrifugal forces caused by the test dummy rotation after the first impact.

The following table shows the maximum acceleration as measured by accelerometer 9 during the 20° slap down (converted to full scale), as well as the acceleration predicted by ADOC.

Measured Versus Predicted Accelerations during Second Impact Accelerometer Location Measured Acceleration Predicted Maximum Acceleration (see Figure 2.10.9-2) (gs) (gs) 61 78 Top (2nd Impact) 9 (Transverse + Rotational) (Transverse + Rotational)

The accelerations measured during the slap down drop test are low relative to the maximum predicted by the ADOC computer program. The acceleration results presented in the above table are taken from the measured acceleration data filtered with a 600 Hz. low pass filter. Figure 2.10.9-20 shows the filtered acceleration time history from accelerometer 9 which experienced the highest acceleration of all locations. The plot of the accelerometer 9 located near impact limiter 4 (second impact, highest acceleration) shows a single peak roughly 0.009 sec. long with a magnitude of approximately 182 g. The initial impact causes a much smaller peak approximately 0.035 sec. prior to the second impact. Note that the acceleration shown in Figure 2.10.9-20 is for the 1/3 scale package, which is equivalent to 3 times the full scale acceleration.

Page 2.10.9-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Crush Depth Measurements After the slap down test, the impact limiter crush depths were measured. There was evidence of both inside and outside crushing. The following table summarizes the measured and predicted crush depths for the top and bottom impact limiters. A spring back of 0.50 inches is assumed (based on previous crush tests). See Appendix 2.10.4 of Reference [3].

Impact Limiter 3 (Bottom)

Maximum Inside Crush Depth (in.) 3.06 Maximum Outside Crush Depth (in.) 1.13 Spring Back (in.) 0.50 Total Crush Depth (in.) 4.69 Predicted Crush Depth 1/3 (in.) 5.17 From the above table it can be seen that the measured crush depth is less than that predicted by the ADOC computer program.

It should also be noted that neither the neutron shield nor the trunnions would contact the target during the impact. Since the crush pattern on the top and bottom impact limiters occur at a 20° angle, and only at the outer edge, there is no possibility of the neutron shield impacting the target during the slap down impact.

Damage Assessment Both impact limiters remained attached to the test dummy during and after the slap down impact.

Considerable crushing from the inside occurred, resulting in significant failure of welds between the impact limiter inner cylinder to the inner base plate. However, since the test dummy remained attached to the limiters, no exposure of the wood to the post-drop fire accident would occur.

Small tears in the impact limiter external shell welds were evident in the impacted area.

However these openings were small and impact limiter wood remained completely confined within the shell.

Figure 2.10.9-21 and Figure 2.10.9-22 are photographs of the test dummy and impact limiters after the 20° slap down drop.

Page 2.10.9-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.9.6.4 90° End Drop Test The forth orientation tested was the 90° end drop. Impact limiters 3 and 4 were reused for this test with impact limiter 3 used for impacting the impact surface. The test article orientation is shown in Figure 2.10.9-6. Two straps were attached to the test articles top two lifting lugs and to the quick release mechanism with padded shackles. Figure 2.10.9-23 is a photograph of the test package set up just before the 90° end drop test.

In order to provide an extreme condition drop test, impact limiter 3 was chilled to -20 °F prior to the drop.

Accelerometer Data The first end drop test results were deemed unacceptable after examination of the accelerometer results. Several of the accelerometer traces showed a sign reversal that could not reflect any physical response of the test dummy. Consequently, the test was redone at a later date using impact limiters 2 and 4 that had been dropped previously.

Impact limiter 2 was used for the impacting limiter because the wood in the end drop crush area was relatively undamaged due to the orientation of the prior drops. However, the second end drop was not done at a chilled temperature. The acceleration time history plots for the second 90° end drop test appeared qualitatively reasonable. The plots generally show a single rounded peak 0.0165 sec. long, with a high frequency low amplitude signal superimposed on top of it. The measured 1/3 scale impact duration of 0.0165 sec. corresponds to 0.0495 sec. for the full size package.

The following table shows the axial acceleration measured by the accelerometers shown, during the 90° end drop, as well as the maximum axial acceleration predicted by ADOC.

Measured Axial Average Axial ADOC Maximum Accelerometer Location Acceleration Acceleration Predicted Axial (see Figure 2.10.9-2) (gs) (gs) Acceleration (gs)

Upper Elevation Avg. 52 CG Elevation Avg. 58 57 49 Bottom Elevation Avg. 60 The acceleration results presented in the above table are taken from the measured acceleration data filtered with a 600 Hz low pass filter. Figure 2.10.9-24 shows the filtered acceleration time history from accelerometer 7. Note that the acceleration plotted in Figure 2.10.9-24 is for the 1/3 scale package, which is equivalent to 3 times the full scale acceleration.

Page 2.10.9-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Crush Depth Measurements After the initial end drop test the crush depths of the bottom impact limiter were measured. There was evidence of both inside and outside crushing. The following table summarizes the measured and predicted crush depths for the bottom impact limiter (impact limiter 3). Note that these results are for the chilled limiter used in the initial end drop. A springback of 0.50 inches is assumed based on Appendix 2.10.4 of Reference

[3].

Impact Limiter 3 Maximum Inside Crush Depth (in.) 2.0 Maximum Outside Crush Depth (in.) 0.50 Spring Back (in.) 0.50 Total Maximum Crush Depth (in.) 3.00 Predicted Total Maximum Crush Depth 1/3 (in.) 3.41 The relatively low crush depth measured after the initial 90° end drop, compared with predicted values can be attributed to the fact that the bottom impact limiter was chilled to -20 °F prior to the drop test.

Damage Assessment Both impact limiters remained attached to the test dummy during and after the first end drop impact, and all impact limiter attachment bolts remained intact.

No openings in the stainless steel impact limiter external shell were evident, and no welds in the external shell failed. Due to the considerable inside deformation, however, the inner disk that contacts the end of the test dummy completely separated from the inner cylindrical shell that encloses the end of the test dummy cylinder. The impact limiter wood remained completely confined within the shell.

Figure 2.10.9-25 and Figure 2.10.9-26 are photographs of the test dummy and impact limiter 3 after the first 90° end drop.

2.10.9.6.5 Puncture Drop Test The final drop test performed was the puncture drop. In order to simulate the proper sequence of accident events specified in 10 CFR 71.73, the impact limiters used for the first end drop test were left on the test dummy without adjustment or tightening of the attachment bolts. Two straps, attached to the top two lifting lugs, were used to support the test model in the same 90° vertical orientation as for the prior end drop. The puncture bar impacted impact limiter 3, which was previously crushed during the 90° end drop. No accelerometer data is provided, since the purpose of the puncture drop is to obtain impact limiter damage data only. Figure 2.10.9-7 depicts the test setup up for the 90° puncture drop test and Figure 2.10.9-27 is a photograph of the test article rigging just prior to the drop.

Page 2.10.9-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Test Results The puncture bar impacted the test package near the center of the outer flat surface of the impact limiter shell. The puncture bar cleanly punched through the outer shell of the impact limiter and was imbedded in the impact limiter wood. The test package came to rest in the vertical position, balanced on top of the puncture bar and then tilted to one side, bending the bar.

The puncture bar sheared a circular section, roughly 2 inches in diameter, of the outer shell of the impact limiter. No other sections of the impact limiter were damaged, and no welds on the impact limiter shell were broken. The impact limiter wood remained completely contained by the impact limiter shell, and no impact limiter wood could be seen at the puncture point.

The puncture bar was stopped by a thin wedge of impact limiter wood that was compacted between the top of the puncture bar and the inner shell of the impact limiter.

The puncture bar did not penetrate the inner stainless steel shell of the impact limiter.

Both impact limiters remained attached to the test dummy during the puncture drop event, and no additional impact limiter attachment bolts were damaged.

Figure 2.10.9-28 and Figure 2.10.9-29 are photographs of the test article after the puncture drop.

2.10.9.7 Conclusions Table 2.10.9-1 summarizes the inertial loads measured during the dynamic testing program. Section 2.7.1 of Chapter 2 goes through detail descriptions to establish the baseline g loads to be used for the cask body, basket, and fuel rod structural evaluations.

The results of the dynamic tests show that:

The crush depths do not result in lockup of the wood in the limiters.

The crush depths for all the drop cases demonstrate that the neutron shield or trunnions would not impact the target.

The impact limiter enclosure is structurally adequate in that it successfully confines the wood inside the steel shell.

The impact limiter attachment design is structurally adequate in that the impact limiters remain on the ends of the test dummy during and after all drop orientations.

A 40 inch drop onto a scaled six inch diameter puncture bar, per 10 CFR 71.73(c)(3), does not significantly damage the impact limiter nor are there any indications of damage to the test dummy. The impact limiters remain firmly secured to the test dummy, and the impact limiter wood is confined.

Page 2.10.9-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.10.9.8 References

1. Title 10, Code of Federal Regulations, Part 71, Packaging and Transportation of Radioactive Material.

2. Mok, Gerald C., et al., Guidelines for Conducting Impact Tests on Shipping Packages for Radioactive Material, UCRL-ID-121673, September, 1995.

3. Transnuclear, Inc., TN-BRP Spent Fuel Package Safety Analysis Report for Transport, Rev. 10, October, 2001, USNRC Docket 71-9202.

Page 2.10.9-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 2.10.9-1 Summary of the Test Results 30 Foot Drop Orientation g Load Measured by Testing 90° End Drop 54 g Axial(1) 0° Side Drop 51 g Transverse(1)

CG Over Corner Drop 34 g Axial 20° Slapdown (Second Impact) 58 g(1)(2), 62 g(1)(3)

Notes:

(1) The g-loads reported in the tables contained in Section 2.10.9.6 of this Appendix are based on the peak value of the raw data recorded from the test results. For design purposes, the region around the peak response has been smoothed to remove the scatter of the test data and provide a representative maximum acceleration. The resulting maximum acceleration values are reported in this table.

(2) The g load measured at this location represents the maximum combined transverse and rotational g load for the basket structural analysis due to the slap down drop case.

(3) The maximum combined g load at the top end of the cask body (at the outer surface of the cask lid).

(4) In order to bound the impact values presented here, the maximum density and minimum moisture content of the full-size impact limiter redwood will be bounded by the values measured for the test articles. The test article average redwood density and moisture content were 23.0 lb/ft3 and 9.8%, respectively. The specified ranges given on SAR drawing 10421-71-41 have been changed to 18.7 - 23.0 lb/ft3 and 9.8 - 15% for the redwood density and moisture content. Balsa values remain unchanged because they have little effect on the impact levels.

Table 2.10.9-2 (deleted)

Page 2.10.9-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-1 One-Third Scale Test Article Page 2.10.9-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-2 Accelerometer Locations Page 2.10.9-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-3 0° Side Drop Test Setup Page 2.10.9-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-4 CG Over Corner Drop Test Setup Page 2.10.9-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-5 20° Slap Down Test Setup Page 2.10.9-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-6 90° End Drop Test Setup Page 2.10.9-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-7 Puncture Drop Test Setup Page 2.10.9-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-8 Test Article and Accelerometer Locations Page 2.10.9-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-9 0° Side Drop Test Rigging Page 2.10.9-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-10 Acceleration Time History, 0° Side Drop Page 2.10.9-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-11 Test Article After 0° Side Drop Page 2.10.9-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-12 Test Article After 0° Side Drop, Upper End Details Page 2.10.9-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-13 Test Article After 0° Side Drop, Lower End Details Page 2.10.9-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-14 TN-40 Impact Limiter Shell Damage After 0° Side Drop Page 2.10.9-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-15 CG Over Corner Drop Test Rigging Page 2.10.9-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-16 CG Over Corner Drop Acceleration Time History Page 2.10.9-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-17 CG Over Corner Drop Impacted Surface Damage Page 2.10.9-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-18 CG Over Corner Drop Impact Limiter Deformation Page 2.10.9-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-19 20° Slap Down Test Rigging Page 2.10.9-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-20 20° Slap Down Drop Acceleration Time History Page 2.10.9-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-21 Test Article After 20° Slap Down Drop Page 2.10.9-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-22 Impact Limiter (Slap Down End) After 20° Slap Down Drop Page 2.10.9-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-23 90° End Drop Test Rigging Page 2.10.9-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-24 Acceleration Time History, 90° End Drop Page 2.10.9-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-25 Test Dummy and Chilled Impact Limiter After Initial 90° End Drop Page 2.10.9-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-26 Test Dummy and Chilled Impact Limiter Post-Drop Detail Page 2.10.9-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-27 Puncture Drop Test Rigging Page 2.10.9-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-28 Test Article After Puncture Drop Page 2.10.9-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 2.10.9-29 Puncture Pin Damage to Impact Limiter Page 2.10.9-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 2.10.10 INNER SHELL BUCKLING DUE TO EXTERNAL PRESSURE TABLE OF CONTENTS 2.10.10.1 Hoop Compressive Stress and Buckling Stresses .......... 2.10.10-1 2.10.10.2 Axial Compressive Stress and Buckling Stresses ........... 2.10.10-2 2.10.10.3 Amplified Axial Stress ...................................................... 2.10.10-3 2.10.10.4 Interaction Equation for Local Buckling ........................... 2.10.10-4 2.10.10.5 References ...................................................................... 2.10.10-4 Page 2.10.10-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 INTRODUCTION The cask containment shell must be shown to withstand the combined external pressure due to fabrication (or shrink fit) and the 290 psi external pressure due to the immersion accident specified in 10CFR71.61. The evaluation provided below follows ASME Code Case N-284 to show that buckling will not occur due to the loads applied.

The gamma shell is conservatively ignored apart from providing the fabrication stress pressure.

2.10.10.1 Hoop Compressive Stress and Buckling Stresses For cylinders of same modulus of elasticity and radial interference , the interface pressure p , is given by [2],

E (b2 a 2 )(c 2 b2 )

p=

2b3 (c 2 a 2 )

Where, a = 36 in (containment cylinder inner radius) [3]

b = 37.5 in (containment cylinder outer radius) [3]

c = 45.5 in (gamma cylinder outer radius) [3]

For a conservative p , a higher E=29.5x106 at 70°F, is assumed [5] for both cylinders.

29.5x10 6 x0.02(37.52 36 2 )(45.52 37.52 )

p=

2x37.53 (45.52 36 2 )

p = 529.02 psi the maximum compressive (hoop) stress due to the shrink-fit pressure occurs in the inner cylinder [2] and is:

2pb 2 2x529.02x37.52 H = 2 = = 13,495 psi 1

(b a 2 ) (37.52 36 2 )

The immersion pressure of 290 psi also produces a compressive hoop stress in the inner cylinder and it is:

pR 290 37.5 H = = = 7,250 psi 3

t 1.5 Combining the two pressures gives the total compressive hoop stress in the inner cylinder:

H = H + H = 13,495 + 7,250 = 20,745 psi T 1 3 Page 2.10.10-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The theoretical buckling stress [4, Sec. 1712] of the inner cylinder depends on whether the cylinder is experiencing hydrostatic (end pressure included) or radial pressure. The shrink fit pressure can be considered a radial pressure while the immersion pressure is a hydrostatic pressure. For this configuration the hydrostatic case bounds (that is, has a lower theoretical buckling stress) and is calculated as R 36.75

= = 24.5 t 1.5 L 169 M = = = 22.76 Rt 36.75x1.5 R

Therefore, 3.0 M 1.65 and t

0.92 0.92 Ch = = = 0.042 M 0.636 22.76 0.636 Et 27.8x10 6 x1.5 eL = heL = Ch = 0.042 = 47,657 psi R 36.75 The combined compressive hoop stress in containment cylinder is evaluated below.

Combined Hoop Stress Accident Condition Calculated Compressive Stress, H 20,745 Factor of Safety, FS [4, Sec. 1400] 1.34 Capacity Reduction, i [4, Sec. 1500] 0.8 H xFS Elastic Amplified Stress, = 34,748 S

i Plastic Reduction Factor, [4, Sec. 1600] 1 Plastic Amplified Stress, S 34,748 Theoretical Buckling Stress, eL 47,657 2.10.10.2 Axial Compressive Stress and Buckling Stresses Axial Compressive Stress due to 290 psig external pressure, pR 290 37.5

= = = 3,625 psi 2t 2 1.5 Theoretical Buckling Stress [4, Sec 1712.1(a)]

Et

= C eL R

Page 2.10.10-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 L

M =

Rt L = 169 in, R avg = 36.75 in, t = 1.5 in, E = 27.8x10 6 psi [5]

169 M =

36.75x1.5 M = 22.76 As, M 1.73 , C = 0.605 (27.8x10 6 x 1.5)

= 0.605 x eL 36.75

= 686,490 psi eL This shows that there is no potential for axial buckling of cylinder.

2.10.10.3 Amplified Axial Stress Calculated axial compressive stress is, = 3,625 psi .

Factor of safety, FS = 1.34 [4, Sec. 1400]

Capacity reduction factor [4, Sec. 1500],

R

1. Effect of t

1.52 0.473 log10 ( i )

= min 1.0 x10 y 0.033 L 5 R 36.75 i = = = 24.5 t 1.5

= 1.52 0.473 log10 (24.5) = 0.863 L

Also, L = 1.0 x 10 -5 x y 0.033 For SA-203, Gr. E, y = 40,000 psi [5]

L = 1.0 x 10 x 40000 0.033 = 0.367

-5 Using smaller value, = 0.367 L

2. Effect of Length

= 0.207 if M 10 L L Page 2.10.10-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Using the larger value of the two [4, Sec 1511],

x FS 3,625 x 1.34 Elastic Amplified Stress, = = = 13,236 psi L

0.367 Plastic amplified stress, = x Elastic Amplified Stres s S

= 1 (Since stress y [4, Sec. 1600])

S = 13,236 psi 2.10.10.4 Interaction Equation for Local Buckling Axial compression plus Hoop Compression

= 13,236 psi , = 34,748 psi , reL = 47,657 psi S S Because < 0.5 and since there is end pressure, heL reL , there is an interaction S S check required because > 0.5 heL [4, Sec. 1713.1.1]

S 2

s 0.5 ha s

+ 1.0 eL 0.5 ha eL 2

13,236 0.5( 47,657 ) 34,748

+ = 0.52 1.0 686,490 0.5( 47,657 ) 47,657 Therefore, the interaction equation is satisfied.

Thus, the inner shell is shown to withstand the combined external pressure due to fabrication stress and the 290 psi immersion accident without buckling.

2.10.10.5 References

1. Not used

2. John Harvey, Theory and Design of Modern Pressure Vessel, Second Edition.

3. TN-40 Transport SAR drawings 10421-1-3, Rev.0.

4. Code Case N-284, Metal Containment Shell Buckling Design Methods,Section III, Division 1, Class MC, August 25, 1980.

5. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices 1989.

Page 2.10.10-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 2.10.11-i and 2.10.11-ii and Pages 2.10.11-1 through 2.10.11-36 Withheld Pursuant to 10 CFR 2.390 Page 2.10.11-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 3 THERMAL EVALUATION TABLE OF CONTENTS Discussion .............................................................................................. 3-1 Summary of Thermal Properties of Materials.......................................... 3-3 Technical Specifications for Components ............................................... 3-8 Thermal Evaluation for Normal Conditions of Transport ......................... 3-8 Thermal Models ........................................................................... 3-8 Maximum Temperatures ............................................................ 3-14 Maximum Accessible Surface Temperature in the Shade .......... 3-14 Minimum Temperatures ............................................................. 3-16 Maximum Internal Pressure ....................................................... 3-16 Maximum Thermal Stresses ...................................................... 3-16 Evaluation of Cask Performance for Normal Conditions of Transport .................................................................................... 3-16 Thermal Evaluation for Hypothetical Accident Conditions..................... 3-19 Fire Accident Evaluation ............................................................ 3-19 Boundary Conditions for the HAC .............................................. 3-19 Crushed Impact Limiter Models.................................................. 3-20 Summary of Results ................................................................... 3-21 Evaluation of Package Performance During and after the HAC Fire .................................................................................... 3-21 References ........................................................................................... 3-22 Appendices ........................................................................................... 3-24 Page 3-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 3-1 NCT Component Temperatures in the TN-40 Package ........................ 3-25 Table 3-2 Temperature Distribution in the TN-40 Package (Low Ambient Temperature, Max Decay Heat) ............................................................ 3-26 Table 3-3 Maximum HAC Transient and Post-Fire Steady-State Maximum Temperatures during Fire Accident ....................................................... 3-27 LIST OF FIGURES Figure 3-1 Schematic of the Cask Body ................................................................. 3-28 Figure 3-2 Thermal Model 90 Degree Radial Cross Section .................................. 3-29 Figure 3-3 Finite Element Model of TN-40 Transport Cask .................................... 3-30 Figure 3-4 Finite Element Model of the TN-40 Transport Cask, Details ................. 3-31 Figure 3-5 Finite Element Model of the TN-40 Basket Cross Section .................... 3-32 Figure 3-6 Finite Element Model of the TN-40 Basket, Details .............................. 3-33 Figure 3-7 Finite Element Model of the TN-40 Basket Compartment Weld Joint Details .......................................................................................... 3-34 Figure 3-8 Details of Impact Limiters...................................................................... 3-35 Figure 3-9 Mesh of Finite Element Model .............................................................. 3-36 Figure 3-10 Temperature Distribution in the TN-40 Cask, NCT, 100 °F ................... 3-37 Figure 3-11 Temperature Distributions in the TN-40 Cask, Fuel & Resin NCT, 100 °F .......................................................................................... 3-38 Figure 3-12 Temperature Distributions in the TN-40 Cask, Impact Limiters &

Rail Normal Conditions Of Transport NCT, 100 °F ............................... 3-39 Figure 3-13 Temperature Distributions In The TN-40 Cask Low Ambient Temperatures ....................................................................................... 3-40 Figure 3-14 Maximum Temperature Distribution in the TN-40 Cask HAC, End of Fire/Smoldering ......................................................................... 3-41 Figure 3-15 Maximum Temperature Distribution in the TN-40 Cask HAC, Cool-Down Period ................................................................................. 3-42 Page 3-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.0 THERMAL EVALUATION Discussion The TN-40 packaging is designed to passively reject decay heat under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) while maintaining packaging temperatures and pressures within specified limits. Objectives of the thermal analyses performed for this evaluation include:

Determination of maximum and minimum temperatures with respect to cask materials limits to ensure components perform their intended safety functions

Determination of temperature distributions to support the calculation of thermal stresses

Determination of the cask cavity gas temperature to support containment pressure calculations

Determination of the maximum fuel cladding temperature Chapter 2 presents the principal design bases for the TN-40 packaging.

The design features of the TN-40 basket are described in Section 1.2. The basket is a welded assembly of stainless steel fuel compartment boxes separated by aluminum and poison plates which form a sandwich panel. The panel consists of two 0.25 in. thick aluminum plates which sandwich a poison plate 0.075 in. thick. The aluminum provides heat conduction paths from the fuel assemblies to the basket peripheral plates. The poison material provides the necessary criticality control. This method of construction forms a very strong honeycomb-like structure of cell liners which provide compartments for 40 fuel assemblies. The aluminum basket rails are bolted to the inner shell and provide a conduction path from the basket to the inner shell. These thermal design features of the basket allow the heat generated by the fuel assemblies to be conducted efficiently from the basket to the shell.

A thermal design feature of the cask is the conduction path created by the aluminum boxes that contain the neutron shielding material as described in Section 1.2. The neutron shielding material is provided by a resin compound cast into long slender aluminum boxes placed around the gamma shield shell and enclosed within a steel outer shell. The aluminum boxes are designed to fit tightly against the steel shell surfaces, thus improving the heat transfer across the neutron shield.

The design of the steel-encased wood impact limiters is described in Section 1.2.

These components are included in the thermal analysis because of their contribution as a thermal insulator. The impact limiters provide protection to the lid and bottom regions from the external heat load applied during the HAC thermal event.

Page 3-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 A personnel barrier prevents access to the outer surfaces of the cask body. The barrier, which consists of a stainless steel mesh attached to stainless steel tubing, will enclose the cask body between the impact limiters, and have an open area of approximately 80%.

Several thermal design criteria are established for the TN-40 to ensure the package meets all its functional and safety requirements. These are:

Containment of radioactive material and gases is a major design requirement. Seal temperatures must be maintained within specified limits to satisfy the NCT leak tight containment requirement. A maximum temperature limit of 536 °F (280 °C) is set for the metallic seals (double metallic O-rings) in the containment vessel lid [11].

To maintain the stability of the neutron shield resin, a maximum allowable NCT temperature of 300 °F (149 °C) is set for the neutron shield [15].

In accordance with 10 CFR 71.43(g) [1] the maximum temperature of accessible package surfaces in the shade is limited to 185 °F (85 °C).

Maximum fuel cladding temperature limits of 400 °C (752 °F) for NCT and 570 °C (1058 °F) for HAC are set for the fuel assemblies with an inert cover gas, as concluded in reference [14].

A temperature limit of 230 °F is set for wood to prevent excessive reduction in structural properties at elevated temperatures [17].

The NCT ambient temperature range is -20 to 100 °F (-29 to 38 °C) per 10 CFR 71.71(b) [1]. In general, all the thermal criteria are associated with maximum temperature limits and not minimum temperatures. All materials can be subjected to the minimum environment temperature of -40 °F (-40 °C) without adverse effects, as required by 10 CFR 71.71 (c)(2) [1].

The TN-40 thermal analysis is conservatively based on a maximum total heat load of 22 kW from 40 fuel assemblies and with a maximum of 0.55 kW per fuel assembly, even though the maximum total heat load is 19 kW. A peak power factor of 1.2 and an active length of 144 in. are considered for calculation of the decay heat profile of the fuel assemblies, as described in Section 3.4.1.3. A description of the detailed NCT analyses is provided in Section 3.4 and HAC analyses in Section 3.5. A summary of the NCT analysis results is provided in Table 3-1. The thermal evaluation concludes that for this thermal design heat load, all design criteria are satisfied.

Page 3-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Summary of Thermal Properties of Materials The analyses use interpolated values when appropriate for intermediate temperatures where the temperature dependency of a specific parameter is significant. The interpolation assumes a linear relationship between the reported values.

Thermal radiation at the external surfaces of the packaging is considered. The thermal analysis assumes that the cask and impact limiter external surfaces are painted white.

Reference [7] gives an emissivity between 0.92 to 0.96 and a solar absorptivity between 0.09 to 0.23 for white paints. To account for dust and dirt and to bound the problem, the thermal analysis uses a solar absorptivity of 0.3 and an emissivity of 0.9 for the white painted surfaces. After a fire, the cask surface will be partially covered in soot (absorptivity = 0.95, Reference [7]). The HAC thermal analysis conservatively assumes an absorptivity of 1.0 and an emissivity of 0.9 for the cool-down period.

1) PWR Fuel Assembly The effective thermal conductivity is calculated using the bounding values, maximum pellet to clad gap and minimum clad thickness, for the PWR fuel assemblies that may be transported in this cask. The fuel conductivity analysis, including the specific heat and density, is presented in Appendix 3.7.1.

Temperature kaxial Temperature ktrans Temperature cP, eff

(°F) (Btu/hr-in-°F) (°F) (Btu/hr-in-°F) (°F) (Btu/lbm-°F) 212 0.0558 136 0.0161 80 0.0593 392 0.0587 233 0.0177 260 0.0654 572 0.0623 330 0.0193 692 0.0726 752 0.0673 428 0.0210 1502 0.0778 932 0.0738 526 0.0228 624 0.0246 eff = 0.135 lb/in3 722 0.0263 821 0.0281 920 0.0298 1019 0.0317

2) 6061 Aluminum (used for basket and rails)

Properties are taken from ASME Section III [3]. The specific heat k

Cp, shown is calculated from: c p =

p where:

p = 0.098 lbm/in.3 [3]

= thermal diffusivity [3]

K = thermal conductivity [3]

Page 3-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Thermal Al 6061 Conductivity, k Specific heat, Cp Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-°F) 70 8.008 0.213 100 8.075 0.215 150 8.167 0.218 200 8.250 0.221 250 8.317 0.223 300 8.383 0.226 350 8.442 0.228 400 8.492 0.230 480 8.492* 0.230*

- value at 400°F conservatively used in this analysis.

3) Poison Plates As a conservative measure, this analysis assumes the Boral plates do not conduct or store heat. A virtual conductivity of 1x10-8 is given to the elements representing Boral in the ANSYS [5] model.

4) Stainless Steel SA 240, Type 304/304L (used for fuel compartments and impact limiter shell) [3]

The stainless steel specific heat is calculated as described above for 6061 Aluminum.

Thermal Conductivity, SA 240, Type 304 k Specific heat, Cp Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-°F) 70 0.717 0.114 100 0.725 0.114 150 0.750 0.117 200 0.775 0.119 250 0.800 0.121 300 0.817 0.122 350 0.842 0.124 400 0.867 0.126 450 0.883 0.127 500 0.908 0.128 550 0.925 0.129 600 0.942 0.130 650 0.967 0.131 700 0.983 0.132 750 1.000 0.132 800 1.017 0.132 850 1.042 0.134 900 1.058 0.134 950 1.075 0.135 Page 3-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Thermal Conductivity, SA 240, Type 304 k Specific heat, Cp Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-°F) 1000 1.100 0.136 1050 1.117 0.136 1100 1.133 0.137 1150 1.150 0.137 1200 1.167 0.138 1431 1.167* 0.138*

= 0.29 lbm/in3 [13]

- value at 1200ºF conservatively used in this analysis.

5) Low Alloy Steel SA 203, Gr E and SA-350, Grade LF3 (containment shell) [3]

Steel specific heat is calculated as described above for 6061 Aluminum.

SA 203 Gr. E or SA Thermal Specific heat , Cp 350 LF3 (3.5 Ni) Conductivity, k Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-°F) 70 1.91 0.107 100 1.93 0.109 200 1.98 0.116 400 1.99 0.127 600 1.91 0.137 800 1.80 0.149 1000 1.68 0.166 1200 1.52 0.203

= 0.284 lbm/in3 [13]

6) Helium (used for gaps within the cask cavity) [12]

Thermal Conductivity, Helium k

Temperature (°F) (Btu/hr-in-°F)

-100 0.0055

-10 0.0064 80 0.0072 260 0.0086 440 0.0102 620 0.0119 980 0.0148 1340 0.0175 For the transient analyses, the thermal mass is relatively small and neglected. The density and specific heat are not used.

Page 3-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

7) SA-516 Grade 70 Carbon Steel (gamma shield shell, outer shell and lid) [3]

Steel specific heat is calculated as described above for 6061 Aluminum.

SA 266, Cl. 4 Thermal SA 516, Gr. 70 or 55 Specific heat, Cp Conductivity, k SA 105 Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-°F) 70 1.97 0.106 200 2.03 0.118 400 2.02 0.128 600 1.93 0.136 800 1.81 0.148 1000 1.67 0.164 1200 1.52 0.188 1400 1.28 0.405

= 0.284 lbm/in3 [13]

8) Air [12]

Thermal Conductivity Air Temperature (ºF) (Btu/hr-in- °F) Prandtl Number

-100 0.0009 0.737 80 0.0013 0.715 260 0.0016 0.705 440 0.0019 0.701 620 0.0022 0.699 980 0.0028 0.701 1340 0.0033 0.710 For the transient analyses, the thermal mass is relatively small and neglected. The density and specific heat are not used.

Page 3-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The following correlations are used to calculate the properties of air to be used in the calculations described in Section 3.4.1.4 Specific Heat Dynamic Viscosity Conductivity (kJ/kg-K) [6] (N-s/m2) [6] (W/m-K) [6]

cP = A(N )T N

= B(N )T N k= C(N )T N

A(0)= 0.103409E+1 For 250 T < 600 K C(0)= -2.276501E-3 A(1)= -0.2848870E-3 B(0)= -9.8601E-1 C(1)= 1.2598485E-4 A(2)= 0.7816818E-6 B(1)= 9.080125E-2 C(2)= -1.4815235E-7 A(3)= -0.4970786E-9 B(2)= -1.17635575E-4 C(3)= 1.73550646E-10 A(4)= 0.1077024E-12 B(3)= 1.2349703E-7 C(4)= -1.066657E-13 B(4)= -5.7971299E-11 C(5)= 2.47663035E-17 For 600 T < 1050 K B(0)= 4.8856745 B(1)= 5.43232E-2 B(2)= -2.4261775E-5 B(3)= 7.9306E-9 B(4)= -1.10398E-12

9) Neutron Shielding (Polyester Resin) [4]

Thermal Conductivity Specific Heat Density (Btu/hr-in- °F) (Btu/lbm- °F) (lb/in3) 0.0083 0.311 0.057

10) Wood Thermal Conductivity (Btu/hr-in- °F)

Min. Max.

0.0019 0.0135 Thermal conductivity values bound values given in Reference [2] for moisture contents up to 30% and specific gravities between 0.08 and 0.80. These values also bound the conductivity parallel and perpendicular to the grain for NCT conditions.

Wood conductivity parallel to the grain is 2.0 to 2.8 times higher than the conductivity across the grain [2]. The maximum wood conductivity, used during the fire accident condition, is taken to be 2.8 times that of the bounding maximum conductivity across the grain to maximize heat conductance from fire toward the cask during fire period.

The maximum wood conductivity during fire is:

K = (2.8)(0.0135 Btu/hr-in-°F) = 0.0378 Btu/hr-in-°F During the transient analyses the thermal mass of the wood is conservatively neglected.

Page 3-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Technical Specifications for Components The cask components for which a thermal technical specification is necessary are the seals.

The seals used in the packaging are the Helicoflex seals (double metallic O-rings). The seals will have a minimum and maximum temperature rating of -40°F and 536°F respectively.

Thermal Evaluation for Normal Conditions of Transport The NCT ambient conditions are used for the determination of the maximum fuel cladding temperature, TN-40 component temperatures, containment pressure and thermal stresses. These steady state environmental conditions correspond to the maximum daily averaged ambient temperature of 100°F and the 10CFR Part 71.71(c)(1) [1] insolation averaged over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

Thermal Models A finite element model is developed using the ANSYS computer code [5]. ANSYS is a comprehensive thermal, structural and fluid flow analysis package. It is a finite element analysis code capable of solving steady-state and transient thermal analysis problems in one, two or three dimensions. Heat transfer via a combination of conduction, radiation and convection can be modeled by ANSYS. The three-dimensional geometry of the cask is modeled. All cask components including the gaps are modeled by SOLID70 conducting elements.

Page 3-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 To determine temperatures of components within the cask body and basket during NCT a finite element model of the basket and cask is developed. The model has 90 degree symmetry and includes the complete cask length. The cask model includes the impact limiters, trunnions, neutron shield, cask shells, cask bottom plate, cask lid, basket, and fuel assemblies (see Figures 3-1 through 3-9). The model simulates the effective thermal properties of the fuel with a homogenized material occupying the volume within the basket where the 144 inch active length of the fuel is stored. The inner shell and gamma shield shell are assembled with an interference fit. This assures thermal contact at the shell interface. The thermal interface conductance increases with contact pressure and decreases with rougher surface finish. At a minimal contact pressure of 5 psi, Reference [15] considers a conductance of 375 Btu/hr-ft2. For conservatism, a 0.01 in. gap with conductivity of 0.0243 Btu/hr-in-°F is considered between the inner shell and the gamma shield shell to represent the interference fit between these shells. A 0.125 in. gap is modeled between the bottom shield and the bottom inner plate. Also, a 0.01 in. axial gap is assumed between the lid outer plate and the shell flange and the lid outer plate and the shield plate. The neutron shielding consists of 60 long resin-filled aluminum containers placed between the gamma shield shell and outer shell. The aluminum containers are confined between these shells, and butt against the adjacent shells. For conservatism, an air gap of 0.01 in. at thermal equilibrium is assumed to be present between the aluminum resin boxes and the adjacent shells. Radiation across these gaps is neglected. Locations of the gaps in the finite element plots of the model are shown in Figure 3-4, Figure 3-6, and Figure 3-7. Figure 3-9 shows the finite element model mesh.

3.4.1.1 Basket Model The basket model is an integrated part of the finite element model which reflects the structure of the basket. The basket structure is composed of 40 stainless steel boxes (8.05 in. square) with two 0.25 in thick aluminum plates and one 0.075 in. thick poison plate placed between adjacent boxes. The boxes are held together by welded stainless steel plugs which pass through the aluminum and poison plates. The plug welding design causes the aluminum and poison plates to be tightly sandwiched between adjacent box sides, (see Figure 3-7. The basket portion of the thermal model simulates the conduction paths provided by the aluminum plates, the stainless steel boxes and the fuel (modeled as a homogenous material). Virtually no thermal conductance is credited to the Boral poison material.

Aluminum plates (0.38 in. thick) are welded to the basket periphery to increase the surface area for heat transfer while providing some structural support for the basket.

These peripheral plates are sized and formed to be in relatively close contact to the inner shell surface at thermal equilibrium. However, the thermal model assumes a 0.10 in gap between these plates and the inner shell surface. The aluminum rails, bolted to the inner shell, are sized so that heat is conducted from the basket periphery across a small gap. All decay heat is transferred from the basket to the inner shell across a helium gap via conduction. Other modes of heat transfer are conservatively neglected.

Figures 3-5 and 3-6 show the basket model with the gaps.

Page 3-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Good surface contact is expected between adjacent components within the basket structure. However to bound the heat conductance uncertainty between adjacent components, the following gaps at thermal equilibrium are assumed:

a. 0.01 in. gap between the weld plugs and adjacent stainless steel fuel compartments.

b. 0.16 in. gap between the small aluminum rail and the basket plate.

c. 0.09 in. gap (modeled as 0.16 in.) between the small conduction plate of the large rail and the basket plate.

d. 0.01 in gap between each adjacent basket plate.

e. 0.10 in gap between the peripheral aluminum basket plates and the inner shell.

The above hot gaps in the model are based on nominal cold gaps shown in the design drawings and are justified as conservative due to the sizes of cold gaps in the design drawing expected to be reduced at thermal equilibrium as discussed in Appendix 3.7.2.

Moreover, the gap sensitivity study based on doubling the largest gap c shows a negligible impact on maximum component temperatures.

The finite element model of the basket includes a representation of the spent nuclear fuel that is based on a fuel effective conductivity model. The decay heat of the fuel with a peaking factor of 1.2 was applied directly to the fuel elements. The maximum fuel temperature reported is based on the results of the temperature distribution in the fuel region of the model. Fuel assembly is modeled as a homogenized material. The effective properties for the homogenized fuel assemblies are described in Appendix 3.7.1.

The gas temperature within the basket is calculated using the temperatures at the hottest cross section of the basket. For simplicity and conservatism, it is assumed to be the average value of the maximum basket plate and cask inner shell temperatures.

3.4.1.2 Impact Limiter Model Similar to the basket model, the impact limiters are an integrated part of the finite element model which determines the maximum wood temperature and the surface temperature of the impact limiters during NCT.

The redwood and balsa within the impact limiters are modeled as a homogenized region containing bounding material properties.

To bound the heat conductance uncertainty between adjacent packaging components the following gaps at thermal equilibrium are assumed:

a. 0.01 in. axial gap between the impact limiter spacer and the cask lid.

b. 0.01 in. axial gap between the impact limiter spacer and the shell flange.

c. 0.0625 in. axial gap between the impact limiter spacer and the top impact limiter.

d. 0.0625 in. axial gap between the bottom shield and the bottom impact limiter.

Page 3-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 All heat transfer across the gaps is by gaseous conduction. Other modes of heat transfer are neglected. The finite element plot of the impact limiter model is shown in Figure 3-8.

3.4.1.3 Decay Heat Load The decay heat load corresponds to a total heat load of 22 kW from 40 assemblies (0.55 kW/assy). A typical heat flux profile for spent PWR fuel with an axial peaking factor of 1.2 was used to distribute the decay heat load in the axial direction within the active fuel length. This heat flux profile is shown below. Within the basket model, the decay heat load is applied as volumetric heat generation in the elements that represent the homogenized fuel.

1.4 Decay Heat Peaking Power 1.2 1

0.8 0.6 0.4 0.2 0

0 10 20 30 40 50 60 70 80 90 100 Percent of Active Fuel Page 3-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.4.1.4 Heat Dissipation Heat is dissipated from the surface of the packaging by a combination of radiation and natural convection.

Heat dissipation by natural convection from various surfaces is described by the following equations using an average Nusselt number [6]:

For horizontal cylinders (cask and impact limiter outer shells):

Nu k hc = with D

D = diameter of the horizontal cylinder k = air conductivity NuT = 0.772 Cl Ra1/ 4 Cl = 0.515 for gases [6]

2f Nu l = Nusselt number for fully laminar heat transfer with ln(1 + 2f / NuT )

0.13 f = 1 (NuT )0.16 Nut = Ct Ra1/ 3 Nusselt number for fully turbulent heat transfer with 1 + 0.0107 Pr Ct = 0.14 for horizontal cylinders [6]

1 + 0.01 Pr Nu = (Nul )m + (Nut )m 1/ m with m = 10 for 1010 Ra 107 g (Tw T ) D 3 Ra = Gr Pr ; Gr =

2 For vertical flat plates (inner and outer cover plates of the impact limiters):

Nu k hc = with L

L = height of the vertical plate k = air conductivity NuT = Cl Ra1/ 4 Cl = 0.515 for gases [6]

2.0 Nu l = Nusselt number for fully laminar heat transfer ln( 1 + 2.0 / Nu T )

Page 3-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Nut = CtV Ra1/ 3 /(1+1.4109Pr/Ra) Nusselt number for fully turbulent heat transfer with 0.13 Pr 0.22 CtV = and f=1+0.078(Tw/T-1).

(1 + 0.61Pr 0.81)0.42 C tV is replaced with f CtV for large (Tw-T).

Nu = (Nul )m + (Nut )m 1/ m with m=6 for 0.1 Ra 1012 g (Tw T ) L3 Ra = Gr Pr ; Gr =

2 Heat transfer from the surface of the package by radiation to the ambient environment is given by (T 14 - T 42 )

h r = F 12 2 o

- Btu/hr - ft - F T1 T 2 Where:

= surface emissivity, F12 = view factor from surface to ambient environment,

= 0.1714 10-8 Btu/hr-ft2-°R4, T1 = surface temperature, °R, and T2 = ambient temperature, °R.

The total heat transfer coefficient Ht = hr+hc, is applied as a boundary condition on the outer surfaces of the finite element model.

3.4.1.5 Solar Heat Load The total insolation for a 12-hour period in a day is 1475 Btu/ft2 for curved surfaces and 738 Btu/ft2 for flat surfaces not transported horizontally as per 10 CFR Part 71.71(c)(1)

[1]. This insolation is averaged over a 24-hr period (daily averaged value) and applied as a constant steady state value to the external surfaces of the thermal models. A solar absorptivity of 0.30 is used for the painted outer surfaces of the packaging. Daily averaging of the solar heat load is justified based on the large thermal inertia of the TN-40 transport package.

Page 3-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Maximum Temperatures Steady state thermal analyses are performed using the maximum decay heat load of 0.55 kW per assembly (22 kW total per cask), 100 °F ambient temperature and the maximum insolation. The temperature distribution within the cask body and basket is shown in Figure 3-10. The temperature distributions as calculated in the fuel assemblies and the neutron shield are shown in Figure 3-11. The temperature distributions within the impact limiter wood and basket rails are shown in Figure 3-12. A summary of the calculated cask component temperatures is listed in Table 3-1.

Maximum Accessible Surface Temperature in the Shade The analysis shows that without the personnel barrier, the maximum accessible cask surface temperature at 100 °F ambient in the shade is 208 °F and exceeds the limit of 185 °F. Therefore, a personnel barrier is required for transport operation at the maximum design basis transportation heat load of 19 kW per cask.

The accessible surfaces of the TN40 packaging consist of the personnel barrier and outermost vertical and radial surfaces of the impact limiters. The personnel barrier surrounds the cask body between impact limiters and has an open area of at least 80%.

The presence of the barrier has negligible effect on heat transfer between the cask surface and the environment. Convection is not affected because the distance between the barrier and the cask and the 80% open area of the barrier ensures that the airflow around the cask is not restricted. Radiant heat transfer to or from the cask surface is not significantly affected because the 80% opening of the barrier allows the cask to see the ambient environment and the distance between the cask and the barrier ensures the screen is very close to ambient temperature. Thus, the 20% of the barrier that the cask sees is also very close to the ambient temperature.

With the installation of the personnel barrier, the accessible packaging surfaces are limited to the impact limiter and the barrier outer surfaces. The NCT model with full insolation at 100 °F ambient temperature shows that the accessible surface temperature of the impact limiters does not exceed 115 °F. The maximum accessible surface temperature of the impact limiters and the maximum cask outer shell temperature at 100 °F ambient in the shade are 106 °F and 208 °F, respectively. The accessible surface temperature of the packaging in the shade is calculated as follows.

The personnel barrier is exposed to thermal radiation from cask outer shell and dissipates heat via thermal radiation and natural convection to ambient. This part of the personnel barrier is over 20 apart from the cask outer shell and is not exposed to the hot air streams from the cask.

Conservatively, omitting the convection heat dissipation from the barrier and assuming the maximum cask outer shell temperature for the entire outer shell surface gives the heat balance for the personnel barrier as follows.

qin,rad = qout,rad Page 3-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 q in,rad =

( 4 Tshell TPB 4

)

1 shell 1 1 PB

+ +

shell Ashell FPBshell APB PB APB q out,rad =

( 4 TPB T 4

)

1 PB 1

+

PB APB FPB APB

= Stefan-Boltzmann constant = 0.119E-10 (Btu/hr-in2-°R4)

Tshell = maximum cask outer shell temperature = 208° F = 668° R TPB = maximum personnel barrier temperature (° R)

T = ambient temperature = 100° F = 560° R shell = emissivity of painted outer shell = 0.9 (see Section 3.2 for discussion)

PB = emissivity of personnel barrier = 0.3 (see Appendix 3.7.1 for discussion)

FPB-shell = view factor from personnel barrier to cask outer shell FPB- = view factor from personnel barrier to ambient = 1.0 APB = surface area of personnel barrier upper part (in2)

Ashell = surface area of cask outer shell upper part (in2)

The diameters of the cask outer shell (Dshell) and personnel barrier (DPB) are 101 and 144, respectively as shown in Drawing 10421-71-2, Rev. 0.

Ashell Dshell 101

APB DPB 144 Since the personnel barrier surrounds the upper part of cask outer shell completely and has an open area of 80%, a view factor of 0.2 can be considered for the cask outer shell to the barrier (Fshell-PB). The view factor of the barrier to the cask outer surface (FPB-shell) can be calculated as follows.

FPBshell APB = FshellPB Ashell Ashell D 101 FPBshell = FshellPB = FshellPB shell = 0.2 = 0.14 APB DPB 144 Substitution of the above values in the heat balance of the personnel barrier gives a maximum temperature of 594° R (134° F).

With the maximum outer surface temperature of the impact limiters at 106 °F and the maximum personnel barrier surface temperature conservatively calculated at 134° F for 100 °F ambient in the shade, the accessible surfaces of the packaging remain below the design criteria of 185 °F (85 °C).

Page 3-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Minimum Temperatures Under the minimum temperature condition of -40 °F (-40 °C) ambient, the resulting packaging component temperatures will approach -40 °F if no credit is taken for the decay heat load. Since the package materials, including containment structures and the seals, continue to function at this temperature, the minimum temperature condition has no adverse effect on the performance of the TN-40.

Temperature distributions at ambient temperatures of -40 °F and -20 °F with maximum decay heat and no insulation are determined. Table 3-2 lists the results of the analyses and the temperature distributions are shown in Figure 3-13.

Maximum Internal Pressure The maximum NCT internal pressure is calculated in Chapter 4.

Maximum Thermal Stresses The maximum NCT thermal stresses are calculated in Chapter 2.

Evaluation of Cask Performance for Normal Conditions of Transport The thermal analysis of NCT demonstrates that the TN-40 cask design meets all applicable requirements, as documented in Table 3-1. The maximum temperatures calculated using conservative assumptions are well below specified limits. The maximum temperature of any containment structural component is less than 251 °F (122 °C). The maximum seal temperature (195 °F, 91 °C) during NCT is well below the 536 °F (280 °C) long-term limit specified for continued seal function. The maximum neutron shield temperature is below 300 °F (149 °C) and no degradation of the neutron shielding is expected. The predicted maximum fuel cladding temperature (495 °F, 257 °C) is well within the allowable fuel temperature limit of 752 °F (400 °C).

The maximum temperature differences across the gaps between various cask layers based on the results of the cask thermal analysis are as follows:

1 °F between the cask inner shell and the gamma shield shell 10 °F between the gamma shield shell and the neutron shield aluminum boxes 12 °F between the neutron shield aluminum boxes and the outer shell The maximum total temperature difference across the cask shells resulting from the gaps is 23 °F.

Although a maximum heat load of 19 kW is allowed during transport of the TN-40 cask, the thermal performance is evaluated for a 22 kW heat load. This conservatism increases the margins of the maximum temperatures to the allowable limits significantly.

Page 3-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Since most of the TN-40 casks to be transported are already loaded and since each cask is limited to a one-time shipment, the following measurements taken prior to shipment ensure the adequacy of the cask thermal performance and compliance with 10 CFR 71.85(a) in lieu of fabrication tests:

Based on the thermal evaluation of the cask, the margins of the fuel cladding and seal temperatures when compared to the allowable limits are:

752 - 495 = 257 °F for fuel cladding, and 536 -195 = 341 °F for seals.

Page 3-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 Page 3-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Thermal Evaluation for Hypothetical Accident Conditions The TN-40 cask is evaluated under the HAC sequence of 10 CFR 71.73 [1]. The top impact limiter protects the TN-40 cask lid containing the lid and port seals from the thermal accident environment. Analytical models are developed as discussed in Sections 3.5.2 and 3.5.3 to demonstrate that seal temperatures are below their material temperature limits during HAC.

Fire Accident Evaluation The fire thermal evaluation is performed primarily to demonstrate the containment integrity of the TN-40. This is assured as long as the metallic seals in the lid remain below 536 °F and the cask cavity pressure is less than 100 psig. A full-length, 90 degree symmetric cask model as described in Section 3.4.1 is used for the evaluation.

The model is modified to represent two crushed impact limiters as described in Section 3.5.3.

Reference [8] reports an average convective heat transfer coefficient of 4.5 Btu/hr-ft2-°F for a railroad tank car fire test. The same parameter is utilized for the HAC fire accident evaluation.

Boundary Conditions for the HAC The boundary conditions described in Section 3.4.1. are modified for the HAC fire.

During the pre-fire and post-fire phases, convection and radiation from the external surface of the model replicate the NCT analysis (100 °F ambient). During the fire phase, a constant convective heat transfer coefficient of 4.5 Btu/hr-ft2-°F is used. All gaps are removed during the fire and restored immediately after the fire. This assumption is conservative in that it ensures maximum heat transfer into the cask during the fire and minimum heat transfer from the cask during the post-fire cooling period. As required by 10 CFR 71.73 [1], a 30 minute 1,475 °F temperature fire with an emittance of 0.9 and a surface absorptivity of 0.8 is applied to the model. An emissivity of 0.9 and an absorptivity of unity are used for the cask external surfaces after the fire accident condition in order to bound the problem.

The sensitivity study that documents the effects of fire emissivity of 1.0 on thermal performance of the TN-40 cask is discussed in Appendix 3.7.3.

A detailed description of the model, including the method used to calculate the maximum fuel cladding temperature and the average cavity gas temperature, is provided in Section 3.4.1. The decay heat load used in this analysis corresponds to a conservative total heat load of 22 kW from 40 assemblies (0.55 kW/assy) with a peaking factor of 1.2 even though the design basis total heat load for transportation condition is 19 kW per cask.

Page 3-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Crushed Impact Limiter Models In order to maximize the effect of the fire on cask components during and after the fire accident, the impact limiter finite element model developed in Section 3.4.1.2 is modified to reflect deformation due to a 30 foot drop. The maximum amount of crush experienced by the impact limiter in a given direction is assumed to occur everywhere on the limiter. Crushed impact limiter configurations based on side, corner and slap down drops are considered:

1. A crushed impact limiter corresponding to the side drop resulting in the shortest radial distance between the fire ambient and the cask surface. The maximum radial deformation of top and bottom impact limiters is 13.42 in. and 13.58 in.,

respectively. The impact limiters are thus modeled with a uniform diameter of 117.2 in. (144-2x13.42) for the top impact limiter and 116.8 in. (144-2x13.58) for the bottom impact limiter to bound the maximum damage at any angular location.

2. A crushed impact limiter corresponding to the corner and end drops resulting in the shortest axial distance between the fire ambient and the cask surface. The maximum axial deformation (17.6 in) occurs for the corner drop. The impact limiters are modeled with a uniform axial length of 20.4 in (38-17.6).

Although the impact limiters are locally deformed during the 30 foot drop, they remain in place on the cask. Under exposure to the thermal accident environment the wood at the periphery of the impact limiter would char but not burn. Hence, the steel encased wood impact limiters still protect the lid of the cask from the external heat load applied during the HAC fire.

Although unlikely, the worst-case damage due to a hypothetical puncture conditions based on 10CFR71.73(c)(3) [1] may result in the tearing off the outer steel skin of the front impact limiter, crushing the wood out of the damaged area, and exposing the partially contained wood to the hypothetical fire conditions.

A study of fire performance of wood at elevated temperatures and heat fluxes [16]

shows that the surface temperature for the rapid spontaneous ignition of wood is between 330 °C and 600 °C (626 °F and 1,112 °F). Based on standard fire test (ASTM E119, 1988) reported in [16], if a thick piece of wood is exposed to fire temperatures between 815 °C and 1,038 °C (1,500 °F and 1,900 °F), the outermost layer of wood is charred. At a depth of 13mm (~0.5) from the active charzone, the wood is only 105 °C (220 °F). This behavior is due to the low conductivity of wood and fire retardant characteristics of char. It is also shown that the char forming rate under high temperature fire conditions is between 37 mm/hr for soft woods and 55 mm/hr for hard woods. Redwood has a char rate of 46 mm/hr [16].

The thickness of Redwood at the center of the TN-40 cask impact limiter is 34.75 inches (883 mm), see Drawing 10421-71-42, Rev. 0, Section 1.4.1. Considering the char rate for Redwood, it takes about 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months until the char reaches 13 mm above the inner surface of the center cover plate. At that moment, the maximum char temperature would be imposed at the impact limiter inner surface.

Page 3-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 (883 13 )

(Redwood thickness - 13) / char rate = = 18 .9 hr 46 It takes another 17 minutes until the last 13 mm of Redwood is charred.

13 (thickness of last portion of hot Redwood) / char rate = = 0.28 hr = 17.0 min 46 During the last 17 minutes, the inner surface of the impact limiter is exposed to the high temperature of charring wood. The impact of charring wood on the cask is maximized if charring occurs immediately after fire for 17 minutes.

To bound the problem and remain conservative, it is considered in the finite element model that the inner surface of the impact limiter inner cover is exposed to 1,112°F maximum char wood temperature for 30 minutes immediately after the end of fire. No heat dissipation is considered for the open surface of the torn wood segment after the period, assuming conservatively that this surface is entirely covered with a thin layer of low conductivity wood char.

Considering the size of wood segments and location of seals, the worst case scenario occurs when a middle segment of wood (ID 44 to OD 88, 90 degree) is torn.

Nevertheless, the effects of a torn side segment (ID 88 OD 144, 30 degree) are also evaluated.

Summary of Results The three investigated accident cases described in Section 3.5.3 show that the maximum component temperatures are bounded by the case of deformed impact limiter with torn middle segment. Table 3-3 presents the bounding maximum temperatures of the cask components during and after the fire event for deformed and torn impact limiter. The bounding values calculated for the maximum seal and the fuel cladding temperatures are 325 °F and 529 °F, respectively. The transient average cavity gas temperature peaks at 387 °F. For conservatism, an average cavity gas temperature of 441 °F is considered for calculating the cavity pressure. The corresponding peak cavity pressure assuming 100% fuel failure is evaluated in Chapter 4 and is less than 100 psig. The temperature distributions in the packaging for HAC are shown in Figure 3-14 and Figure 3-15.

Evaluation of Package Performance During and after the HAC Fire It is concluded that the TN-40 maintains containment during the postulated sequential drop, puncture, and fire accident. The neutron shield will off-gas during the fire event.

A pressure relief valve is provided on the outer shell to prevent the pressurization of the outer shell. The shielding integrity of the neutron shielding is assumed to be lost after the fire event and the resulting accident dose rates have been evaluated in Chapter 5.

The maximum seal temperature is well below the 536 °F long-term limit specified for continued seal function and the fuel cladding temperature is below the limit of 1058 °F (570 °C) defined in [14].

Page 3-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. Code of Federal Regulations, 10CFR71, Packaging and Transportation of Radioactive Materials.

2. Wood Handbook: Wood as an Engineering Material, U.S. Department of Agriculture, Forest Service, March 1999.

3. ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers,Section III and Appendices, 1989.

4. TN-24 Dry Storage Cask Topical Report, Transnuclear, Inc., Revision 2A, Hawthorne, NY, 1989.

5. ANSYS Engineering Analysis System, User's Manual for ANSYS, Release 8.0 and 8.1 ANSYS, Inc., Canonsburg, PA.

6. Rohsenhow, M. W., Cho, Y. I., and Harnett, J. P., Handbook of Heat Transfer, 3rd Edition, 1998.

7. Siegel, Howell, Thermal Radiation Heat Transfer, 4th Edition, 2002.

8. Gregory et al., Thermal Measurements in a Series of Large Pool Fires, SAND85-0196, TTC-0659, Sandia National Laboratories, 1987.

9. NUREG/CR-0497, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, MATPRO, Version 11 (Revision 2),

EG&G Idaho, Inc., TREE-1280, August 1981.

10. Baumeister & Marks, Standard Handbook for Mechanical Engineers, Seventh Edition, McGraw-Hill Book Co., New York, 1969.

11. Helicoflex High Performance Sealing Catalog, Carbone Lorraine, Helicoflex Components Division, ET 507 E 5930.

12. W. M. Rohsenow, J. P. Hartnett, Handbook of Heat Transfer Fundamentals, 2nd Edition, 1985.

13. Perry, Chilton, Chemical Engineers Handbook, 5th Edition, McGaw-Hill, Inc. New York, 1973.

14. USNRC, SFPO, Cladding Consideration for the Transportation and Storage of Spent Fuel, Interim Staff Guidance ISG-11, Rev. 3.

15. Prairie Island independent Spent Fuel Storage Installation Safety Analysis Report, Revision 10.

Page 3-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

16. Mitchell S. Sweet, Fire Performance of Wood: Test Methods and Fire Retardant Treatments, Fire Safety of Wood Products, USDA Forest Service.

http://www.fpl.fs.fed.us/documents/pdf1993/sweet93a.pdf.

17. NUREG/CR-0322, Effects of Temperature on the Energy Absorbing Characteristics of Redwood, Von Riesemann, Gues, SAND77-1589, Sandia Lab.

18. Transnuclear, Inc., Letter to USNRC, TN-32 Cask Thermal Testing, Docket No. 72-1021, December 1, 2000, Transnuclear Document No. E-18578, Project 1066.

19. EPRI, High-Burnup Used Fuel Dry Storage System Thermal Benchmark Modeling Results, Round Robin Results, Final Report, April 2020.

20. North Anna Power Station ISFSI - Amendment No.5 to Materials License No.2507 for the North Anna Power Station Independent Spent Fuel Storage Installation (CAC No. L25047), NRC Accession Number: ML17234A534.

21. High Burnup Dry Storage Research Project Cask Loading and Initial Results, EPRI 3002015076, October 2019.

22. North Anna Power Station ISFSI - TN-32 DLBD, Revision 8, HBU Demonstration Cask Design/Licensing Basic Document, NRC Accession Number: ML17109A457.

Page 3-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Appendices Appendix 3.7.1 Effective Thermal Properties for the Fuel Assembly Appendix 3.7.2 Justification of Hot Gap between Basket and Cask Inner Shell Appendix 3.7.3 Sensitivity Study for Effects of Fire Emissivity Page 3-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 3-1 NCT Component Temperatures in the TN-40 Package Normal Transport Maximum Minimum* Allowable Component (°F) (°F) Range(°F)

Outer Shell 214 -40 **

Radial Neutron Shield 229 -40 -40 to 300 Inner Shell 251 -40 **

Basket Rail 257 -40 **

Basket (Fuel Compartments) 444 -40 **

Gamma Shield Shell 248 -40 **

Fuel Cladding 495 -40 752 max.

Impact Limiter Wood 224 -40 230 Cask Bottom Inner Plate 234 -40 **

Cask Lid 192 -40 **

Vent and Drain Port Seal***** 192 -40 -40 to 536 Lid O-ring Seal**** 195 -40 -40 to 536 Average Cavity Gas*** 345 -40 N/A Accessible Surface Temperature in Shade 134 -40 185 max

Assuming no credit for decay heat and a daily average ambient temperature of -40 °F

- The components perform their intended safety function within the operating range.

- - A conservative value of 348 °F is used for calculating MNOP.

- - - The elements between the cask inner shell and cask lid at radius (cylindrical x-coordinate) between 36.43 and 41.38 and height (cylindrical z-coordinate) between 164.55 and 171.55 represent the location of the lid seal in the model.

- - - - The elements within the cask lid at radius (cylindrical x-coordinate) between 22.2 and 37.5 and height (cylindrical z-coordinate) between 171.49 and 173.70 represent the location of the vent and drain port seal in the model.

Page 3-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 3-2 Temperature Distribution in the TN-40 Package (Low Ambient Temperature, Max Decay Heat)

Maximum Component Temperature ( °F)

-40 °F Ambient -20 °F Ambient Component Temperature Temperature Outer Shell 88 106 Radial Neutron Shield 102 120 Gamma Shield Shell 123 140 Inner Shell 127 144 Basket Rails 133 151 Fuel Cladding 386 401 Cask Bottom Inner Plate 109 126 Cask Lid 63 81 Vent and Drain Port Seal** 63 81 Lid O-ring Seal* 68 86 Basket (Fuel Compartments) 330 346

The elements between the cask inner shell and cask lid at radius (cylindrical x-coordinate) between 36.43 and 41.38 and height (cylindrical z-coordinate) between 164.55 and 171.55 represent the location of the lid seal in the model.

- The elements within the cask lid at radius (cylindrical x-coordinate) between 22.2 and 37.5 and height (cylindrical z-coordinate) between 171.49 and 173.70 represent the location of the vent and drain port seal in the model.

Page 3-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 3-3 Maximum HAC Transient and Post-Fire Steady-State Maximum Temperatures during Fire Accident Maximum Post-Fire Steady-State Maximum Transient Temperature Allowable Component Temperature (°F) (°F)***** Range (°F)

Impact Limiter Outer Surface 1431 147 **

(end of fire)

Outer Shell Surface 1084 252 **

(end of fire)

Cask Bottom Inner Plate 343 260 **

(One hour after fire)

Cask Lid 289 231 **

(4.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after fire)

Vent and Drain Port Seal**** 284 230 536 (4.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after fire)

Lid O-ring Seal*** 325 229 536 (one hour after fire)

Gamma Shield Shell 694 273 **

(end of fire)

Cask Rail / Shim 330 283 **

(one hour after fire)

Inner Shell 403 277 **

(end of fire)

Basket (Fuel Compartment) 480 474 **

(20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire)

Fuel Cladding 529 524 1058 (26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months after fire)

Average Cavity Gas 387* 374 **

(10.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after fire)

An average cavity gas temperature of 441 °F is considered for calculating of cavity gas pressure in Chapter 4 for conservatism.

- The components perform their intended safety function within the operating range.

- - The elements between the cask inner shell and cask lid at radius (cylindrical x-coordinate) between 36.43 and 41.38 and height (cylindrical z-coordinate) between 164.55 and 171.55 represent the location of the lid seal in the model.

- - - The elements within the cask lid at radius (cylindrical x-coordinate) between 22.2 and 37.5 and height (cylindrical z-coordinate) between 171.49 and 173.70 represent the location of the vent and drain port seal in the model.

- - - - Thermal analysis results at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after end of fire conservatively used.

Page 3-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-1 Schematic of the Cask Body Page 3-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-2 Thermal Model 90 Degree Radial Cross Section Page 3-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Top Impact Limiter Neutron Shield Wood Basket Upper Trunnion Lower Trunnion Bottom Impact Limiter Figure 3-3 Finite Element Model of TN-40 Transport Cask Page 3-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-4 Finite Element Model of the TN-40 Transport Cask, Details Page 3-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-5 Finite Element Model of the TN-40 Basket Cross Section Page 3-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-6 Finite Element Model of the TN-40 Basket, Details Page 3-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-7 Finite Element Model of the TN-40 Basket Compartment Weld Joint Details Page 3-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

1. Center ring and inner gussets are not included in the model for normal transport conditions.

These features are considered only for accident conditions.

Figure 3-8 Details of Impact Limiters Page 3-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-9 Mesh of Finite Element Model Page 3-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-10 Temperature Distribution in the TN-40 Cask, NCT, 100 °F Page 3-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-11 Temperature Distributions in the TN-40 Cask, Fuel & Resin NCT, 100 °F Page 3-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-12 Temperature Distributions in the TN-40 Cask, Impact Limiters & Rail Normal Conditions Of Transport NCT, 100 °F Page 3-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-13 Temperature Distributions In The TN-40 Cask Low Ambient Temperatures Page 3-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-14 Maximum Temperature Distribution in the TN-40 Cask HAC, End of Fire/Smoldering Page 3-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3-15 Maximum Temperature Distribution in the TN-40 Cask HAC, Cool-Down Period Page 3-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 3.7.1 EFFECTIVE THERMAL PROPERTIES FOR THE FUEL ASSEMBLY TABLE OF CONTENTS 3.7.1.1 Discussion ........................................................................... 3.7.1-1 3.7.1.2 14 x 14 PWR Fuel Geometry Parameters ........................... 3.7.1-1 3.7.1.3 Summary of Material Properties .......................................... 3.7.1-1 3.7.1.4 Thermal Model .................................................................... 3.7.1-3 3.7.1.5 Effective Density and Specific Heat ..................................... 3.7.1-9 3.7.1.6 Conclusion......................................................................... 3.7.1-10 3.7.1.7 References ........................................................................ 3.7.1-11 LIST OF FIGURES Figure 3.7.1-1 Finite Element Model Of WE 14 X 14 Assembly (1/4 Symmetry) .............................................................................. 3.7.1-12 Figure 3.7.1-2 Typical Boundary Conditions Based On FE Model Of WE 14 X 14 ................................................................................... 3.7.1-13 Figure 3.7.1-3 Typical Temperature Distributions For FE Model Of WE 14 X 14 ........................................................................................ 3.7.1-14 Figure 3.7.1-4 Comparison of Transverse Effective Conductivities ................ 3.7.1-15 Figure 3.7.1-5 Comparison of Time Temperature Histories ........................... 3.7.1-16 Page 3.7.1-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 EFFECTIVE THERMAL PROPERTIES FOR THE FUEL ASSEMBLY 3.7.1.1 Discussion In order to determine the effective fuel assembly thermal conductivity, effective fuel density, and effective specific heat, the 14x14 PWR fuel assemblies to be transported in the TN-40 cask were reviewed to select the fuel assembly or parameters that would provide the most conservative (lowest) effective thermal conductivity. Use of these properties would conservatively predict bounding maximum temperatures for the TN-40 cask.

Effective conductivity values in the axial and transverse directions are calculated separately. The transverse fuel effective conductivity is determined by creating a two-dimensional finite element model of the fuel assembly centered within a basket compartment using the ANSYS computer code [5]. The outer surfaces, representing the fuel compartment walls, are held at a constant temperature, and a decay heat is applied to the fuel pellets within the model. A steady state solution of the model determines the maximum fuel assembly temperature. The two-dimensional model is described in Section 3.7.1.4.

3.7.1.2 14 x 14 PWR Fuel Geometry Parameters The fuel assemblies to be transported in the TN-40 cask are WE14x14 Standard, WE14x14 OFA, Exxon Standard, Exxon High Burnup, and Exxon TOPROD.

The bounding values for the gap between the pellet and the clad (maximum gap of 0.0048 in.) and for the clad thickness (minimum of 0.0225 in.) are used for this evaluation. These values give the minimum transverse effective conductivity.

The material properties used to calculate the effective fuel properties are listed in Section 3.7.1.3.

3.7.1.3 Summary of Material Properties

a. UO2 [3]

Temperature (C) k (cal/s-cm-C) Temperature (F) k (Btu/hr-in-F) 25 0.025 77 0.503 100 0.021 212 0.423 200 0.018 392 0.362 300 0.015 572 0.302 500 0.0132 932 0.266 700 0.0123 1292 0.248 800 0.0124 1472 0.250 Page 3.7.1-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Temperature (C) Cp (cal/g-C) Temperature (F) Cp (Btu/lbm-F) 0 0.056 32 0.056 100 0.063 212 0.063 200 0.0675 392 0.068 400 0.0722 752 0.072 1200 0.079 2192 0.079 The density of fuel pellets (UO2) is 10.96 g/cc = 0.396 lbm/in3 [3]

b. Zircaloy-4 [2]

Temperature (K) k (W/m-K) Temperature (F) k (Btu/hr-in-F) 373.2 13.6 212 0.655 473.2 14.3 392 0.689 573.2 15.2 572 0.732 673.2 16.4 752 0.790 773.2 18.0 932 0.867 873.2 20.1 1112 0.968 Temperature (K) Cp (J/kg-K) Temperature (F) Cp (Btu/lbm-F) 300 281 80 0.067 400 302 260 0.072 640 331 692 0.079 1090 375 1502 0.090 The density of Zircaloy is 6.56 g/cm3 = 0.237 lbm/in3, as defined in Reference [3].

Table B-3.11 of Reference [2] lists the measured emissivity values for fuel cladding.

For ease of calculation a temperature independent emissivity of 0.80 is set for zircaloy-4.

zirc = 0.80

c. Helium (used for gaps within the cask cavity) [6]

Helium Thermal Conductivity, k Temperature (F) (Btu/hr-in-F)

-100 0.0055

-10 0.0064 80 0.0072 260 0.0087 440 0.0102 620 0.0119 980 0.0148 1340 0.0175 Page 3.7.1-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 For the transient analyses, the thermal mass is relatively small and neglected. The density and specific heat are not used.

d. Stainless Steel A hemispherical emissivity of 0.46 is reported in [7] for 304 stainless steel samples.

For conservatism an emissivity of 0.3 is considered in this calculation for the link elements representing the stainless compartment walls to create the radiation superelement.

3.7.1.4 Thermal Model 3.7.1.4.1 Transverse Effective Conductivity The purpose of the effective conductivity in the transverse direction of a fuel assembly is to relate the temperature drop of a homogeneous heat generating square to the temperature drop across an actual assembly cross section for a given heat load. The isotropic effective thermal conductivity of a heat generating square, such as the fuel assembly, can be calculated from the following equation from [4]:

q a 2 k trans = (0.29468)

(Tc To )

(1) q = volumetric heat generation rate (Btu/hr-in.3) a = half of the compartment width = 8.05 in./2 = 4.025 (in.)

Tc = maximum center temperature (peak cladding temperature) (°F)

To = wall temperature (°F)

The volumetric heat generation rate is:

Q q =

4 a 2 La (2)

Q = decay heat load per assembly = 0.675 kW 1 = 2,303.3 Btu/hr La = active fuel length = 144 (in.)

Substituting equation (2) in (1) gives:

Q k trans = (0.29468 )

4La (Tc To )

(3)

In determining the temperature dependent effective fuel conductivities an average temperature, equal to (Tc + To)/2, is used for the fuel temperature.

1 0.675 kW is the decay heat load per assembly for the TN-40 storage cask [1]. For transportation, maximum heat load per assembly in TN-40 cask is limited to 0.55 kW.The effective transverse conductivity is relatively insensitive to decay heat load.

Page 3.7.1-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.7.1.4.2 Finite Element Model A discrete finite element model of the fuel assembly in the TN-40 fuel compartment is developed using the ANSYS computer code [5]. This two-dimensional quarter-symmetry model of the fuel assembly simulates heat transfer by radiation and conduction and includes the geometry of the fuel rods and guide tubes. Helium is used as the fill gas in the fuel assembly. A fuel assembly decay heat load of 0.675 kW is used for heat generation. An active length of 144 in. (366 cm) is assumed. Radiation between the fuel pellet and cladding is conservatively neglected. The fuel assembly is centered within the fuel compartment.

The fuel components were modeled using PLANE55 elements. No convection is considered within the fuel assembly model. Heat transfer from the fuel rods to the fuel compartment walls is through conduction and radiation.

Radiation between the fuel rods, the guide tubes, and the fuel compartment walls was simulated using the radiation super-element processor (/AUX12). LINK32 elements were used to define radiating surfaces to create the radiation super-element. The LINK32 elements were unselected prior to the solution of the model. The model was run with a series of isothermal boundary conditions applied to the outermost nodes representing fuel compartment walls.

LINK32 elements were located at symmetry axes to make an enclosure for creating the radiation super-element. A very low emissivity (0.001) is given to the LINK32 elements lying on the symmetry axes to minimize their effect on overall radiation heat transfer.

The conductivity of helium is considered for the back fill gas for transport conditions.

The thermal properties used are as described in Section 3.7.1.3, and the fuel assembly geometry is described in Section 3.7.1.2. Figure 3.7.1-1 shows the details of the finite element model. A typical boundary condition is shown in Figure 3.7.1-2. A typical temperature distribution is shown in Figure 3.7.1-3.

The maximum fuel assembly temperatures resulting from the 2D analysis are shown in the table below. The reaction solution (Qreact) equals the decay heat load per active length of the fuel assembly. Since the model is quarter-symmetric, the applied decay heat load is:

Q = 4 Qreact La (4)

Substituting equation (4) in equation (3) gives the effective fuel conductivity in the transverse direction.

Qreact k trans = (0.29468)

(Tc To )

Page 3.7.1-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Several computational runs were made using isothermal boundary temperatures ranging from 100 to 1000ºF. The results of the finite element analysis and the effective fuel assembly conductivity calculation using the equation above are:

T0 Tc Tavg Qreact(1) ktrans (F) (F) (F) (Btu/hr-in) (Btu/hr-in-F) 100 173 136 3.9753 0.0161 200 266 233 3.9753 0.0177 300 361 330 3.9753 0.0193 400 456 428 3.9753 0.0210 500 551 526 3.9753 0.0228 600 648 624 3.9753 0.0246 700 744 722 3.9752 0.0263 800 842 821 3.9753 0.0281 900 939 920 3.9753 0.0298 1000 1037 1019 3.9754 0.0317 Q (Btu/hr) / kW 2290/0.671(2)

Notes:

(1) The negligible value difference has no effect on ktrans.

(2) Small difference with 0.675 kW calculated in Section 3.7.1.4.1 has no effect on ktrans.

A sensitivity study is performed to capture the effect of UO2 conductivity on the fuel assembly effective transverse conductivity. The following UO2 properties are considered in this sensitivity analysis.

Temperature kUO2 (Btu/hr-in-F) kUO2 (Btu/hr-in-F)

(F) (SCALE Module) (1) (MATPRO) (2) 77 0.503 0.398 212 0.423 0.345 392 0.362 0.294 572 0.302 0.256 932 0.266 0.203 1292 0.248 0.169 1472 0.250 0.156 Notes:

(1) Based on SCALE Modules from [3]

(2) Based on MATPRO from [2] (95% theoretical density (TD) UO2). The equation (A.2-1.a) from MATPRO is used to calculate fuel pellet thermal conductivity for 0C<T1650C.

Fuel assembly transverse effective conductivities based on the above fuel pellet (UO2) thermal conductivities are calculated using the same methodology described above.

The results are shown in the following table and compared in Figure 3.7.1-4 to the transverse effective conductivities used in thermal analysis of TN-40 cask.

Page 3.7.1-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Based on KUO2 from [3] Based on KUO2 from [2]

Tavg ktrans Tavg ktrans (F) (Btu/hr-in-F) (F) (Btu/hr-in-F) 130 0.0193 130 0.0193 226 0.0224 226 0.0223 323 0.0258 323 0.0257 420 0.0297 420 0.0296 517 0.0343 517 0.0341 615 0.0395 615 0.0392 713 0.0450 713 0.0446 811 0.0512 812 0.0507 910 0.0580 910 0.0573 1009 0.0651 1009 0.0646 As seen from Figure 3.7.1-4, transverse effective conductivity using fuel pellet thermal conductivity from MATPRO is 1.5% lower than the values obtained using fuel pellet thermal conductivity from SCALE. However, both calculated values of transverse effective conductivity are at least 20% higher than those reported in Section 3.2.1. The primary reason for this large difference is due to use of a Stefan-Boltzmann constant of 1.983E-13 Btu/hr-in2-R4 in the original calculation. This value is 60 times lower than the correct value of 1.190E-11 Btu/hr-in2-R4.

Since the transverse effective conductivities from Section 3.2.1 are lower than the corrected values assigned as Based on UO2 from NUREG/CR-0200 Rev. 6 (Scale) shown in Figure 3.7.1-4, the results for NCT discussed in Section 3.4.2 were not changed and considered as conservative.

A sensitivity analysis described below demonstrates that the transverse effective conductivities of the fuel assembly used in the transient analysis of HAC are also conservative and the results shown for HAC in Section 3.5.4 remain unchanged.

The ANSYS model described in Sections 3.5.1 to 3.5.3 is modified in this sensitivity analysis as follows. The transverse effective conductivities of the fuel assemblies are changed to the corrected values based on MATPRO data and the fire emissivity is changed to 1.0. All other thermal properties and boundary conditions are the same as those described in Sections 3.5.1 through 3.5.3. The modified model is run with boundary conditions of NCT and subsequently with boundary conditions of HAC to capture the effects of the effective transverse conductivity on the peak component temperatures.

The peak component temperatures from the modified model for HAC are compared to those reported in Table 3.7.3-1 in the following table.

Page 3.7.1-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Allowable Steady State Limit Transient Temperatures Cool-Down Temperatures (1) (°F) f=1.0 f=1.0 (with increased keff (with increased for Fuel f=1.0 keff for Fuel f=1.0 Assemblies) [SAR Table 3.7.3-1] Assemblies) [SAR Table 3.7.3-1]

Tmax Tmax Tmax Tmax Tmax Tmax Component (°F) (°F) (°F) (°F) (°F) (°F) 513 531 Fuel cladding -18 507 526 -19 1058 (5)

(20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire) (26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months after fire)

Basket 480 482

-2 473 476 -3 (2)

(fuel compartment) (20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire) (20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire) 339 339 Cask rail/shim 0 285 284 +1 (2)

(0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after fire) (0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after fire) 420 420 Cask inner shell (4) 0 278 278 0 (2)

(end of fire) (end of fire)

Gamma shield 726 726 0 274 274 0 (2) shell (end of fire) (end of fire) 333 333 Lid O-ring seal(3) 0 230 230 0 536 (5)

(1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after fire) (1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after fire) 1138 1138 Cask outer shell 0 253 253 0 (2)

(end of fire) (end of fire)

Impact limiter 1477 1478

-1 147 147 0 (2) surface (end of fire) (end of fire) 385 389 Average cavity gas (13.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after (13.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after -4 371 375 -4 ---

fire) fire)

(1) Thermal analysis results at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the end of fire is used conservatively to bound the steady state temperatures.

Actual steady state temperatures in both runs will achieve the same values.

(2) The components perform their intended safety function within the operating range.

(3) The elements between the cask shell flange and cask lid at radius (cylindrical x-coordinate) between 36.43 and 41.38 and height (cylindrical z-coordinate) between 164.55 and 171.55 represent the location of the lid seal in the model.

(4) Includes shell flange.

(5) See Table 3.7.3-1 for the references of the allowable limits.

Page 3.7.1-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 As seen in the above table, the effect of increasing the transverse effective fuel conductivity results in a decrease of -18°F for the peak fuel cladding temperature for HAC. The peak fuel cladding temperature with the increased effective fuel conductivity is 513°F and occurs approximately at 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the end of the fire. This time is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months shorter than the 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months at which the peak fuel cladding temperature of 531°F occurs for fuel assemblies with lower effective fuel conductivity. This is due to faster heat-up rate of the system caused by the increased effective fuel conductivity. However, the maximum temperatures remain lower due to lower initial temperatures.

As seen in the above table, for steady state conditions after cool-down period the increased effective fuel conductivity has negligible effect on the maximum temperature of the cask components while reduces the maximum temperature of the fuel cladding significantly.

The time-temperature histories of the maximum component temperatures for the cask outer shell and fuel cladding from the above sensitivity study and the sensitivity study in Appendix 3.7.3 are compared together in Figure 3.7.1-5.

Based on the peak temperature results and comparisons presented in Figure 3.7.1-5, the peak temperatures of all components are bounded by those reported in Table 3-3.

This demonstrates that using the transverse effective fuel conductivities as reported in Section 3.2.1 are conservative in evaluation of thermal performance of TN-40 transport cask.

3.7.1.4.3 Axial Effective Conductivity The axial fuel assembly conductivity can be calculated by taking credit for the conduction paths provided by the fuel cladding, the guide tube and the helium in the fuel compartment. However, in accordance with [8], the axial effective conductivity calculated here is limited to the conductivity of the cladding. The axial conductivity provided by the fuel pellets conservatively neglected.

The effective conductivity is determined by weighting the cladding conductivity by its fractional area to the fuel compartment area:

cladding area k axial = cladding conductivity 4a 2 a = half of compartment width = 8.05/2 = 4.025 Page 3.7.1-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Zr-4 Cladding Area (AZr) is calculated as:

The results are summarized in Section 3.7.1.6.

3.7.1.5 Effective Density and Specific Heat Volume average density and weight average specific heat are calculated to determine the effective density and specific heat for the fuel assembly. The equations to determine the effective density and specific heat are shown below.

eff =

V i i

=

UO 2 VUO 2 + Zr 4 VZr 4 Vassembly 4a 2La C p,eff =

V Ci i Pi

=

UO 2 VUO 2 CP ,UO 2 + Zr 4 VZr 4 CP ,Zr 4 V i i UO 2 VUO 2 + Zr 4 VZr 4 The results are summarized in Section 3.7.1.6.

Page 3.7.1-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.7.1.6 Conclusion The minimum temperature-dependent effective conductivities, density, and specific heat have been calculated for the 14 x 14 PWR fuel to be transported in the TN-40 cask and are summarized below:

Axial Transverse Specific Conductivity, Conductivity, Heat, Temperature kaxial Temperature ktrans Temperature cP, eff (F) (Btu/hr-in-F) (F) (Btu/hr-in-F) (F) (Btu/lb-F) 212 0.0558 136 0.0161 80 0.0593 392 0.0587 233 0.0177 260 0.0654 572 0.0623 330 0.0193 692 0.0726 752 0.0673 428 0.0210 1502 0.0778 932 0.0738 526 0.0228 624 0.0246 eff = 0.135 lb/in3 722 0.0263 821 0.0281 920 0.0298 1019 0.0317 Page 3.7.1-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.7.1.7 References

1. Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Revision 10.

2. NUREG/CR-0497, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, MATPRO - Version 11 (Revision 2),

EG&G Idaho, Inc., TREE-1280, August 1981.

3. Oak Ridge National Laboratory, RSIC Computer Code Collection, SCALE, A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation for Workstations and Personal Computers, NUREG/CR-0200, Rev. 6, ORNL/NUREG/CSD-2/V3/R6.

4. SANDIA Report, SAND90-2406, A Method for Determining the Spent Fuel Contribution to Transport Cask Containment Requirements, 1992.

5. ANSYS, Inc., ANSYS Engineering Analysis System Users Manual for ANSYS Revision 8.0 and 8.1, Cannonsburg, PA.

6. W.M. Rohsenow, J.P. Harnett, Handbook of Heat Transfer Fundmentals, 2nd Edition, 1985.

7. Azzazy Technology Inc., Emissivity Measurements of 304 Stainless Steel, Report Number ATI-2000-09-601, 2000.

8. NUREG-1536, Standard Review Plan for Dry Cask Storage Systems - Final Report, 1997.

Page 3.7.1-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3.7.1-1 Finite Element Model Of WE 14 X 14 Assembly (1/4 Symmetry)

Page 3.7.1-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Heat generating Boundary Conditions Constant Temperature at Compartment Wall Figure 3.7.1-2 Typical Boundary Conditions Based On FE Model Of WE 14 X 14 Page 3.7.1-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 3.7.1-3 Typical Temperature Distributions For FE Model Of WE 14 X 14 Page 3.7.1-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 0.07 UO2 from NUREG/CR-0200 Rev.6 UO2 from NUREG/CR-0497 0.06 Section 3.2.1 0.05 K_eff (Btu/hr-in- oF) 0.04 0.03 0.02 0.01 0 200 400 600 800 1000 1200 T_avg ( oF)

Figure 3.7.1-4 Comparison of Transverse Effective Conductivities Page 3.7.1-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Fuel Cladding 550 530 Temperature ( oF) 510 490 Fire Emissivity =1.0 Fire Emissivity =0.9 Increased Fuel keff, Fire Emissivity =1 470 450 0 5 10 15 20 25 30 35 40 45 Time (hr)

Cask Outer Shell 1400 1200 Fire Emissivity =1.0 1000 Fire Emissivity =0.9 Temperature ( oF)

Increased keff, Fire Emissivity

=1 800 600 400 200 0

0 5 10 15 20 25 30 35 40 45 Time (hr)

Figure 3.7.1-5 Comparison of Time Temperature Histories Page 3.7.1-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 3.7.2 JUSTIFICATION OF HOT GAP BETWEEN BASKET AND CASK INNER SHELL TABLE OF CONTENTS 3.7.2.1 Discussion ........................................................................... 3.7.2-1 3.7.2.2 References .......................................................................... 3.7.2-2 LIST OF TABLES Table 3.7.2-1 Average Temperatures at Hottest Cross Section...................... 3.7.2-3 Table 3.7.2-2 Average Coefficient of thermal Expansions .............................. 3.7.2-3 Page 3.7.2-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 JUSTIFICATION OF HOT GAP BETWEEN BASKET AND CASK INNER SHELL 3.7.2.1 Discussion The inner diameter of the cask inner shell is 72 +00..25 00 inches based on Drawing 10421 3, Rev. 0. This gives a nominal diameter of 72.125 for cask inner shell. The maximum outer diameter of the basket is 71.92 based on Drawing 10421-71-9, Rev. 0.

Considering a tolerance of 0.25 for the basket gives a nominal basket diameter of 71.795.

To calculate the minimum gap, the average temperatures for the basket aluminum plates and cask inner shell at the hottest cross section for NCT at 100F ambient are retrieved from the finite element model. The hottest cross section is located at z=60.31 in the model where the maximum fuel cladding temperature is achieved. The average temperatures are listed in Table 3.7.2-1.

The average coefficient of thermal expansions for the aluminum 6016 and SA-203, Gr.

D or E are listed in Table 3.7.2-2 The hot dimensions of the basket OD and cask inner shell ID are calculated as follows.

The outer diameter of the hot basket is:

ODB,hot = ODB [1+ Al6061 (Tavg,B - Tref)] = 72.078 Where:

ODB,hot = hot OD of the basket (in)

ODB = nominal cold OD of the basket = 71.795 Al6061 = Average thermal expansion coefficient for aluminum 6061 at Tavg,B =

13.41E-6 in/in-F (interpolated using data from Table 3.7.2-2)

Tavg,B = Average basket aluminum plates temperature at the hottest cross section =

364 F, see Table 3.7.2-1 Tref = reference temperature = 70 F [1]

The inner diameter of the hot cask inner shell is:

IDcask,hot = IDcask [1 + SA203 (T avg,cask - Tref)] = 72.208 Where:

IDcask,hot = hot ID of the cask inner shell (in)

IDcask = nominal cold ID of the cask inner shell = 72.125 SA203 = Average thermal expansion coefficient for SA-203, Gr. D or E at Tavg,cask =

6.65E-6 in/in-F (interpolated using data from Table 3.7.2-2)

Page 3.7.2-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Tavg,cask = Average cask inner shell temperature at the hottest cross section = 244 F, see Table 3.7.2-1 Tref = reference temperature = 70 F [1]

The radial hot gap between the basket and cask inner shell is:

Gaphot = (IDcask,hot - ODB,hot) = (72.208 - 72.078) / 2 = 0.065 A uniform radial hot gap of 0.10 is considered in the model between the basket aluminum plates and the cask inner shell. This assumption is conservative since the calculated hot gap of 0.065 is smaller than the assumed gap of 0.1.

3.7.2.2 References

1. ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers,Section II, 1989.

Page 3.7.2-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 3.7.2-1 Average Temperatures at Hottest Cross Section Component Material Tavg (F)

Basket aluminum plates Al-6061 364 Cask inner shell SA-203, Gr. D or E 244 Table 3.7.2-2 Average Coefficient of thermal Expansions Temperature (F) Al-6061 [1] SA-203, Gr. D or E [1]

(in/in-F) (in/in-F) 200 12.91E-6 6.54E-6 300 13.22E-6 6.78E-6 400 13.52E-6 6.98E-6 Page 3.7.2-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 3.7.3 SENSITIVITY STUDY FOR EFFECTS OF FIRE EMISSIVITY TABLE OF CONTENTS 3.7.3.1 Discussion ........................................................................... 3.7.3-1 3.7.3.2 Conclusion........................................................................... 3.7.3-2 3.7.3.3 References .......................................................................... 3.7.3-2 LIST OF TABLES Table 3.7.3-1 Maximum Component Temperature for 22kW Heat Load for Fire Emissivity =1.0 and 0.9 ................................................ 3.7.3-3 LIST OF FIGURES Figure 3.7.3-1 Results of the Sensitivity Study................................................. 3.7.3-4 Page 3.7.3-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 SENSITIVITY STUDY FOR EFFECTS OF FIRE EMISSIVITY 3.7.3.1 Discussion A fire emissivity of 0.9 was used in Section 3.5.2 to calculate the fire radiation heat transfer to the TN-40 cask during the fire. Assuming conservatively, the fire as a black body, an emissivity of 1.0 can be used for the fire. The effect of this assumption is determined for the TN-40 cask in a sensitivity analysis in this section. According to Section 3.5.4, the case of a deformed impact limiter with a torn middle segment represents the bounding accident case and is chosen for the sensitivity analysis. The methodology and assumptions used in the sensitivity analysis are the same as those described in Section 3.5 for the HAC thermal evaluation except for the increase of the fire emissivity from 0.9 to 1.0.

A comparison of the maximum TN-40 component temperatures based on fire emissivities of 0.9 and 1.0 is shown in Table 3.7.3-1.

As seen from Table 3.7.3-1, the effect of increasing the fire emissivity from 0.9 to 1.0 on maximum fuel cladding temperature is an increase of 2F. The predicted maximum fuel cladding temperature of 531 F (277 C) is well within the allowable fuel cladding temperature limit of 1058 F (570 C) [1], [2] for accident conditions.

The largest effect of increasing the fire emissivity from 0.9 to 1.0 occurs at the cask outer shell when it is directly exposed to the fire. The other components remain shielded from the fire so that the inner shell temperature increases by only 17 F and the cask rail temperature increases by only 9 F.

These temperature increases are relatively small and last for a short period of time and therefore do not affect the thermal and structural performance of the TN40 transport cask.

The containment seals are protected from direct fire exposure by the impact limiters.

The effect of increasing the fire emissivity from 0.9 to 1.0 on the maximum seal temperatures is limited to 8 F for a short period of time after the fire. The transient and the steady state temperatures of the containment seals remain well below the temperature limit of 536 F (280C) [3]. Therefore the containment function of the seals is unaffected by the increase of the fire emissivity from 0.9 to 1.0.

The time-temperature histories of the maximum component temperatures for the cask outer shell and fuel cladding from the sensitivity study with fire emissivity of 1.0 compared to those from the original model with fire emissivity of 0.9 are shown in Figure 3.7.3-1.

Page 3.7.3-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 As seen in Figure 3.7.3-1, the TN-40 cask component temperatures decrease through the cool-down period. The small differences seen in Table 3.7.3-1 between the steady state temperatures are caused by the fact that the transient temperatures at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after fire are picked to bound the steady state temperatures.

3.7.3.2 Conclusion The effect of increasing the fire emissivity from 0.9 to 1.0 lasts only for a short period of time on the outermost components of the cask exposed to fire. The function of the other cask component remains unaffected by this change in the fire emissivity.

3.7.3.3 References

1. Code of Federal Regulations, 10CFR71, Packaging and Transportation of Radioactive Materials.

2. USNRC, SFPO, Cladding Consideration for the Transportation and Storage of Spent Fuel, Interim Staff Guidance ISG-11, Rev. 3.

3. Helicoflex High Performance Sealing Catalog, Carbone Lorraine, Helicoflex Components Division, ET 507 E 5930.

Page 3.7.3-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 3.7.3-1 Maximum Component Temperature for 22kW Heat Load for Fire Emissivity =1.0 and 0.9 Steady State Transient Temperature (F) Temperature (1) f=1.0 f=0.9 f=1.0 f=0.9 Allowable Tmax Tmax Tmax Tmax Tmax Tmax Limit Component (°F) (°F) (°F) (°F) (°F) (°F) (°F) 531 529 Fuel cladding +2 526 524 +2 (26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months after fire) (26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months after fire) 1058 [2]

Basket (fuel 482 480

+2 476 474 +2 (2) compartment) (20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire) (20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after fire) 339 330 Cask rail/shim +9 284 283 +1 (2)

(0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after fire) (1.0 hour0 days 0 hours 0 weeks 0 months after fire) 420 403 Cask inner shell(4) +17 278 277 +1 (2)

(end of fire) (end of fire) 726 694 Gamma shield shell +32 274 273 +1 (2)

(end of fire) (end of fire) 333 325 Lid O-ring seal(3) +8 230 229 +1 (1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after fire) (1.0 hour0 days 0 hours 0 weeks 0 months after fire) 536 [3]

1138 1084 Cask outer shell +54 253 252 +1 (2)

(end of fire) (end of fire) 1478 1431 Impact limiter surface +47 147 147 0 (2)

(end of fire) (end of fire) 389 386 Average cavity gas +3 375 374 +1 -

(13.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after fire) (10.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after fire)

(1) Thermal analysis results at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the end of fire is used conservatively to bound the steady state temperatures. Actual steady state temperatures in both runs will achieve the same values.

(2) The components perform their intended safety function within the operating range.

(3) The elements between the cask shell flange and cask lid at radius (cylindrical x-coordinate) between 36.43 and 41.38 and height (cylindrical z-coordinate) between 164.55 and 171.55 represent the location of the lid seal in the model.

(4) Includes shell flange.

Page 3.7.3-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Fuel Cladding Fire Emissivity =1.0 Fire Emissivity =0.9 550 F) o 530 Temperature (

510 490 470 450 0 5 10 15 20 25 30 35 40 45 Time (hr)

Cask Outer Shell Fire Emissivity =1.0 Fire Emissivity =0.9 1400 F) 1200 o

1000 Temperature (

800 600 400 200 0

0 5 10 15 20 25 30 35 40 45 Time (hr)

Figure 3.7.3-1 Results of the Sensitivity Study Page 3.7.3-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 4 CONTAINMENT TABLE OF CONTENTS Containment Boundary ........................................................................... 4-1 Containment Vessel ..................................................................... 4-1 Containment Penetrations ............................................................ 4-1 Seals and Welds .......................................................................... 4-2 Closure ......................................................................................... 4-2 Requirements For Normal Conditions Of Transport ................................ 4-3 Containment of Radioactive Material ........................................... 4-3 Pressurization of Containment Vessel ......................................... 4-8 Containment Criterion ................................................................ 4-10 Containment Requirements for Hypothetical Accident Conditions ............................................................................................. 4-10 Source Terms............................................................................. 4-11 Containment of Radioactive Material ......................................... 4-11 Containment Criterion ................................................................ 4-11 Leakage Rate Tests for Type B Packages............................................ 4-13 Special Requirements ........................................................................... 4-14 References ........................................................................................... 4-14 Page 4-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 4-1 Radionuclide Inventory and A2 Values .................................................. 4-15 Table 4-2 Activity Concentration by Source .......................................................... 4-16 Table 4-3 Normal Conditions of Transport and Hypothetical Accident Conditions Effective A2 Values ............................................................. 4-17 Table 4-4 Normal Condition of Transport and Hypothetical Accident Conditions Permissible Leakage Rates from the TN-40 ....................... 4-18 Table 4-5 Total Moles of Fission Gas for a WE 14x14 Std. Assembly .................. 4-19 Table 4-6 Cask Gas Mixtures under Normal Conditions of Transport and Hypothetical Accident Conditions ......................................................... 4-20 LIST OF FIGURES Figure 4-1 TN-40 Containment Boundary Components ......................................... 4-21 Figure 4-2 Lid, Vent Port And Drain Port Metal Seals ............................................ 4-22 Page 4-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 4.0 CONTAINMENT Containment Boundary The containment boundary components consist of the inner shell and bottom inner plate, shell flange, lid outer plate, and vent / drain port covers. Also, included are the associated seals and bolts. The containment boundary is shown in Figure 4-1. The construction of the containment boundary is shown on Drawings 10421-71-3, 4 and 5 provided in Appendix 1.4. The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. Helium assists in heat removal and provides a non-reactive environment to protect fuel assemblies against fuel cladding degradation which might otherwise lead to gross rupture.

Containment Vessel The TN-40 containment vessel consists of an inner shell which is a welded carbon steel cylinder and is welded to a carbon steel bottom inner plate and a shell flange forging.

The vessel closure is a carbon steel lid with bolts, vent cover with bolts, and drainport cover with bolts. The lid outer plate thickness is 4.5 in. The overall containment vessel length is 170.5 in. with a wall thickness of 1.5 in. The cylindrical cask cavity has a diameter of 72.0 in. and a length of 163 in.

The containment shell and bottom inner plate materials are SA-203 Grade D or E and the shell flange is SA-350 Grade LF3. The lid outer plate material is SA-203 Grade E or SA-350 Grade LF3.

The cask design, fabrication and testing are performed under Transnuclear's Quality Assurance Program which conforms to the criteria in Subpart H of 10CFR71 [2].

The materials of construction meet the requirements of Section III, Subsection NB-2000 and Section II, Material Specifications [3] or the corresponding ASTM Specifications.

The containment vessel is designed to the ASME Code,Section III, Subsection NB, Article 3200 to the maximum practicable extent. The containment vessel is fabricated and examined in accordance with NB-2500, NB-4000 and NB-5000. Also, weld materials conform to NB-2400 and the material specification requirements of Section II, Part C of ASME B&PV. The containment vessel is hydrostatically tested in accordance with the requirements of the ASME B&PV Code,Section III, Article NB-6200.

Alternatives to the ASME Code are specified in SAR Section 2.11.

Containment Penetrations There are two penetrations through the containment vessel, both in the lid. One is the drain port and the other is the vent port. A double seal mechanical closure is provided for each penetration. Each penetration incorporates a bolted cover.

Page 4-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Seals and Welds The containment boundary welds consist of the circumferential welds attaching the bottom inner plate and the shell flange to the inner shell. Also, the longitudinal weld(s) on the rolled plate, closing the cylindrical inner shell, and the circumferential weld(s) attaching the rolled shells together are containment welds.

Double metallic seals (O-rings) are utilized on the lid and the two lid penetrations.

Helicoflex HND [1] or equivalent seals may be used. The seals are shown in Figure 4-2. The Helicoflex metallic face seals of the lid and lid penetrations possess long-term stability and have high corrosion resistance. These high performance metallic seals consist of an inner spring, a lining, and a jacket. The spring is Nimonic 90 or equivalent material. The lining and jacket are stainless steel and aluminum, respectively.

Additionally, all metallic seal seating surfaces are stainless steel for improved surface control.

The internal spring and lining maintain the necessary rigidity and sealing force, and provide some elastic recovery capability. The outer aluminum jacket provides a ductile material that ensures leak tightness. The jacket also provides a connecting sheet between the inner and outer seals. Holes in this sheet allow for attachment screws and for communication between the overpressure port (OP) cover and the space between the seals. This sheet, which is approximately 0.020 in. thick, has insufficient strength to transmit radial forces great enough to overcome the axial compressive forces on the seals. The OP seal is a single metallic seal of the same design, Helicoflex HN200 or equivalent.

The lid and penetration seals described above are contained in grooves in the lid or port covers. A high level of sealing over the transport period is assured by utilizing seals in a deformation-controlled design. The deformation of the seals is constant since bolt loads assure that the mating surfaces remain in contact. The seal deformation is set by the original O-ring cross section and the depth of the groove. The specified preload has the required force to seat the seals as calculated in Appendix 2.10.2.2.3.

The spring is made of Nimonic 90 or equivalent material that has equivalent or better corrosion properties, service temperature and modulus of rigidity and elasticity which ensures the seal will not be affected by relaxation and thus the seal can be maintained at the specified temperatures for extended periods.

Helicoflex metallic seals (Reference [1]) are all capable of limiting leak rates to less than 1 x 10-7 ref cm3/sec. After loading for transport, all lid and cover seals are leak tested in accordance with ANSI N14.5. The acceptable total cask leakage (both inner and outer seals combined) is 1 x 10-4 ref cm3/sec.

Closure The containment vessel contains an integrally-welded bottom closure and a bolted and flanged top closure (lid). The outer lid plate is attached to the shell flange with 48 bolts.

Page 4-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The bolt torque required to seal the metallic seals located in the lid and maintain containment under normal and accident conditions is provided in Drawing 10421-71-1 provided in Appendix 1.4. The closure bolt analysis is presented in Appendix 2.10.2.

As previously mentioned, the lid contains two penetrations which are sealed by flanged cover plates fastened to the lid by 8 bolts each. The bolt torque required to seal the metallic seals in the penetration covers and maintain containment under normal and accident conditions is provided in Drawing 10421-71-1 provided in Appendix 1.4.

Requirements For Normal Conditions Of Transport In accordance with 10 CFR 71.51, a Type B package must be designed, constructed and prepared for shipment so that no loss or dispersal of radioactive contents, as demonstrated to a sensitivity of 10-6 A2 per hour will occur under the tests specified in 10 CFR 71.71 for normal conditions of transport.

The guidelines of ANSI N14.5 [6] were used to determine the leakage test criteria which demonstrate that the TN-40 meets the no-loss requirements of 10 CFR 71.51.

Containment of Radioactive Material 4.2.1.1 Source Terms Three sources are considered to determine the releasable airborne material from the TN-40 cask [4].

Residual activity on the cask interior surfaces as a result of loading operations (and, if applicable, previous shipments);

Fission and activation-product activity associated with corrosion-deposited material (crud) on the fuel assembly surfaces, and Radionuclides within the individual fuel rods comprising the fuel assemblies.

The first source, residual contamination on the interior surfaces of the cask is neglected.

Reference [4] indicates that this is negligible as compared to the crud deposition on the fuel rods.

The second source, crud, is basically the radioactive flaky material that is formed on the outside surface of the fuel rods due to the radioactive and corrosive environment of the PWR reactor. This material can be loosely bound to the fuel rod surface and may be dislodged during transportation and be available for release from the cask.

Page 4-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The third source is from the fuel itself. A breach in the fuel cladding may allow radionuclides to be released from the fuel to the interior of the cask. There are three types of radionuclides releases associated with the breaches in the fuel rod cladding:

gaseous, volatiles and fuel fines.

For conservatism, it is assumed that crud spallation and cladding breaches occur instantaneously after fuel loading and closure operations. Therefore, all radioactivity is readily available for release if a leak occurs.

The containment analysis is based on the void volume within the TN-40 cask.

The cavity volume is 6.64E+05 in3 (volume of a 72 in. x 163 in. cylinder). From SAR Table 2-6, the basket contains 6,610 lb of steel and 8,080 lb of aluminum. Using a density of 0.29 lb/in3 for steel and 0.098 lb/in3 for aluminum, the basket volume is calculated as 1.05E+05 in3. With a fuel rod OD of 0.426 in. and a rod length of 152 in.,

the volume for 40 fuel assemblies each containing 179 fuel rods is calculated to be 1.55 E+05 in3. The volume for the fuel assembly hardware (spacer grids, guide tubes, and instrument tube) is approximately 218 in3 per assembly. Thus the calculated fuel assembly volume (40 assemblies) is 1.64E+05 in3. Calculating the cask void volume:

Cask Void Volume = 6.64E+05 in.3 - 1.05E+05 in.3 - 1.64E+05 in.3

= 3.95E+05 in.3

= 6.46E+06 cm3 Source Activity from the Fuel The fuel transported in the TN-40 transport packaging has a maximum assembly average initial enrichment of 3.85 wt% U-235, 45,000 MWD/MTU bundle average exposure and a minimum of 15 - 25 year cooling time provided the fuel acceptance criteria of Section 1.2.3 have been met. As discussed in Chapter 5, Section 5.2, numerous SAS2H [8] evaluations were performed to determine the design basis fuel for shielding. These SAS2H analyses were also evaluated to determine the bounding fuel parameters for the containment analysis. The bounding SAS2H evaluation was performed for the 14 x 14 Westinghouse standard fuel assembly with 39,000 MWD/MTU burnup, enrichment of 3.3 wt. % U-235 and a cooling time of 15 years. It is assumed that 40 design basis fuel assemblies are loaded in the TN-40 cask transport packaging. The radionuclide inventory consists of activity from iodine, fission products that contribute greater than 0.1% of the design basis fuel activity and actinides that contribute greater than 0.01% of the design basis activity. Tritium is also included although it contributes slightly less than 0.1% of the design basis activity. The radionuclide inventory is presented in Table 4-1.

Source Activity from Release of Volatiles The source activity concentration inside the TN-40 due to the release of volatiles, Cvolatiles, is calculated using the following equation [4].

Page 4-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Cvolatiles = {NAfBAVfv} / V where:

NA = number of assemblies, fB = fraction of rods that develop cladding breaches, AV = specific activity of volatiles in the fuel assembly, Ci/assembly, fV = fraction of volatiles in a fuel rod released if a fuel rod develops a cladding breach, and V= void volume inside the containment vessel, cm3 Table 4-2 presents the results of this calculation.

Source Activity from Release of Gaseous Isotopes The source activity concentration inside the TN-40 cask due to the release of gaseous isotopes, Cgases, is calculated using the following equation [4].

Cgases = {NAfBAGfG} / V where:

NA = number of assemblies, fB = fraction of rod that develop cladding breaches, AG = specific activity of gases in the fuel assembly, Ci/assembly, fG = fraction of gases in a fuel rod released if a fuel rod develops a cladding breach, and V= void volume inside the containment vessel, cm3.

Table 4-2 presents the results of this calculation.

Source Activity from Release of Fuel Fines The source activity concentration inside the TN-40 due to the release of fuel fines, Cfines, is calculated using the following equation [4].

Cfines = { NA fB AF fF } / V where:

NA = number of assemblies, fB = fraction of rod that develop cladding breaches, AF = specific activity of fuel fines in the assembly, Ci/assembly, ff = fraction of fuel fines released if a fuel rod develops a cladding breach, and V = void volume inside the containment vessel, cm3.

Table 4-2 presents the calculated concentration of fuel fines inside the TN-40 for normal transport conditions.

Page 4-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Source Activity due to Crud Spallation The fuel transported in the TN-40 transport packaging may be cooled a minimum of 15 to 25 years (provided the fuel acceptance criteria of Section 1.2.3 have been met). The activity density that results inside of the TN-40 as a result of crud spallation, Ccrud, is calculated using the equation below [4].

Ccrud = { fC SC NR NA SAR } e-t / V where:

fC = crud spallation factor, V = free volume inside the containment vessel, cm3, SC = crud surface activity, Ci/cm2, NR = number of fuel rods per assembly, NA = number of assemblies in the cask, SAR = surface area per rod, cm2, and e-t = decay factor ( = 0.693/5.27 and t = 15 yr).

The surface area of the 14 x 14 fuel rods calculated for this containment analysis is presented below.

SAR = ( d l) + 1/4(2 d2) where:

d = rod outer diameter = 0.422 in = 1.07 cm (from Table 5-3), and l = rod length = 152 in = 386.1 cm (from Table 5-3) substituting and solving:

SAR = 1.30E+03 cm2 / rod 4.2.1.2 Determination of A2 Values The A2 value of a mixture of radioactive nuclides is determined as follows:

A2 mixture = [ (fi / A2i )] -1 where:

fi is the fraction of total activity due to isotope i, and A2i is the A2 value for isotope i.

Using the methodology of 10 CFR 71 and Reference [4], the A2 values are determined for each source (Table 4-1). The data provided in Table 4-1 (radionuclide inventory and A2 values) and Table 4-2 (source activity) are combined to determine an effective A2 for the TN-40 (Table 4-3).

Page 4-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 4.2.1.3 Determination of Permissible Leakage Rates To determine the leakage rates, the four sources are combined to form the total source term:

Ctotal = Ccrud + Cvolatiles + Cgases + Cfines From Reference [6], the permissible release rate, R, from the TN-40 is:

R=LC where:

L = volumetric gas leakage rate (cm3/s),

C = curies per unit volume of the radioactive material that passes through the leak path, and R = A2 x 2.78 x 10-10 /second for normal transport conditions.

For normal conditions, the permissible leakage rate is 1.50E-04 cm3/sec (Table 4-4).

This value is converted to units of ref-cm3/sec by first calculating the equivalent hole size. From ANSI N14.5 [6]:

Lu = upstream volumetric leakage rate, cc/sec = 1.50E-04 cm3/sec Fc = coefficient of continuum flow conductance per unit pressure, cm3/atm-sec Fm = coefficient of free molecular flow conductance per unit pressure, cm3/atm-sec Pu = fluid upstream pressure, atm abs = 2.50 atm abs (conservative value)

Pd = fluid downstream pressure, atm abs = 1.0 atm abs D = leakage hole diameter, cm a = leakage hole length, cm = 0.5 cm (assuming leak path length is on the order of the metal seal width)

= fluid viscosity, cP = 0.028 cP T = fluid absolute temperature, 222C = 495 K (average cavity gas temperature a conservative value based on Table 3-1)

M = molecular weight, g/g-mol = 4 g/g-mol (from ANSI N14.5, Table B.1)

Pa = average stream pressure = 1/2 (Pu + Pd), atm abs = 1.75 atm abs Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec where:

Fc = (2.49x106 D4)/(a) cm3/atm-sec, and Fm = {3.81x103 D3 (T/M)0.5} / {aPa} cm3/atm-sec.

Page 4-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Substituting:

Fc = (2.49x106 D4)/(0.5 0.028) = 1.78E+08 D4 Fm = {3.81x103 D3 (495/4)0.5} / {0.5 1.75} = 4.84E+04 D3 Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec 1.50E-04 = (Fc + Fm) (2.50 - 1.0) (1.75 / 2.50) 1.50E-04 = (Fc + Fm) (1.05)

Fc + Fm = 1.43E-04 Solving the equations above for D, yields a hole diameter of 8.87 x 10-4 cm.

This equivalent hole size, is then used to calculate the reference air rate at standard conditions. Assuming all upstream test conditions correspond to standard conditions:

Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec where:

Pu = air upstream pressure, atm abs = 1.00 atm abs Pd = air downstream pressure, atm abs = 0.01 atm abs

= air viscosity, cP = 0.0185 cP (from ANSI N14.5, Table B.1)

T = air ref temperature = 298° K M = molecular weight air, g/g-mol = 29 g/g-mol (from ANSI N14.5, Table B.1)

Pa = average stream pressure = 1/2 (Pu + Pd), atm abs = 0.505 atm Substituting:

Fc = {2.49E+06 (8.87E-04) 4}/(0.5 0.0185) = 1.66E-04 Fm = {3.81E+03 (8.87E-04)3 (298/29.0)0.5} / {0.5 0.505} = 3.37E-05 Lstd = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec Lstd = (1.66E-04 + 3.37E-05)(1.0 - 0.01)(0.505 / 1.0)

Lstd = 1.0E-04 ref cm3/s Pressurization of Containment Vessel The TN-40 cask cavity is drained, dried and evacuated prior to backfilling with helium at the end of fuel loading operations. If the TN-40 cask contains design basis fuel and has been in storage for a short period prior to shipment (i.e., thermal equilibrium is reached),

the cask cavity temperature with 100F ambient air and maximum solar load is 401F.

The maximum normal operating pressure during storage is 2.2 atm abs [5].

Similarly, during normal transport conditions, the maximum cavity gas temperature is 348F (176C) under hot environment conditions. The maximum initial cavity pressure just prior to shipment (assuming no fuel rod failure) is 2.0 atm abs. The operational procedure guidelines for conducting these activities are provided in Chapter 7.

Page 4-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Cavity Gas Mixtures The determination of fission gases is based on the grams of fission gases from SAS2H /

ORIGEN -S computer runs [8], which utilizes fuel with 45,000 MWD/MTU bundle average exposure, 3.8 U-235 wt% initial bundle average enrichment and 15 year cooled. The gases which are considered following irradiation are: iodine (I), krypton (Kr), and xenon (Xe). The bulk of the fission gases remain trapped in the fuel pellet.

The release fraction of 0.3 is applied to these gases. Table 4-5 presents the total moles of fission gas for the fuel assembly.

In addition to the fission gasses, gas due to helium from rod pre-pressurization is included. From Reference [5] the free gas volume after irradiation is 9.04 m3/cask at standard temperature and pressure. Therefore the mass of free gas can be determined as:

n = VP/RT

= 9.04 m3 x (1 atm x (1 bar/0.9869 atm)) / ((0.08314 bar m3/kg mole °K X 273°K)

= 4.036E-01 kg mole/cask.

The mass of free He gas (pre-pressurization), therefore, can be determined by subtracting the irradiation generated gas (Table 4-5) from the total free gas:

0.4036 - 0.3119 = 0.0917 Kg mole/cask.

The third and final component of gas is helium from cask backfilling operations which is estimated from the ideal gas law below and shown in Table 4-6:

n = VP / RT

= 6.46 m3/cask x (2.2 atm x (1 bar/0.9869 atm)) / ((0.08314 bar m3/kg mole °K) x 449°K)

= 3.858E-01 kg mole/cask.

The cavity gas mixture (assuming 3% fuel rod failure) is 97.7% helium (from cask backfill operations and from rod pre-pressurization) with the balance consisting of xenon (2.0%), krypton (0.2%), iodine (0.1%). These results are presented in Table 4-6. This gas mixture is not explosive.

Page 4-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Maximum Normal Operating Pressure (MNOP)

The mechanisms contributing to containment pressurization are ideal gas heating and release of fission gas from the fuel rods. The maximum normal operating pressure is calculated using the recommendation of NUREG-1617 [7] which uses the following conditions:

30% release rate of fission gas from fuel pellets into the gap between the fuel pellets and the cladding.

3% failure rate of fuel rod cladding.

maximum cavity gas temperature of 348F (176C) under hot environment conditions.

the gas volume (plenum and pellet to cladding volume) inside the fuel rods is conservatively neglected when calculating the cask free volume.

The fuel assemblies release a total of 0.0121 Kg-mole (9.35E-03+2.756E-03, Table 4-6). The pressures are calculated below:

P3% rod failure = (0.0121Kg-moles/cask 0.08314 bar m3/Kg mole °K 449°K)/(6.46 m3)

P3% rod failure = 0.070 bars = 0.069 atm abs MNOP = Pinitial + P3% rod failure MNOP = 2.0 atm abs + 0.069 atm abs = 2.07 atm abs = 30.4 psia Therefore, the maximum normal operating pressure for the TN-40 is 30.4 psia (15.7 psig). Casks designs with MNOP greater than 5.0 psig must be subjected to a structural pressure test in accordance with 10 CFR 71.85(b). The test pressure must be at least 1.5 times MNOP. The TN-40 will be subjected to a hydrostatic test at a pressure of 25 psig. This test is described in Chapter 8.

Containment Criterion As will be demonstrated in Section 4.3.3, the reference leak rate for normal conditions, 1.0E-04 ref cm3/s, is a significantly lower rate than the accident leakage rate. The acceptance criterion for the fabrication, periodic verification, and pre-shipment leakage test shall be set at 1.0E-04 ref cm3/s.

Containment Requirements for Hypothetical Accident Conditions The containment requirement under hypothetical accident conditions are specified by 10 CFR 71.51(a)(2). It states there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week.

It is assumed for purposes of the accident condition evaluation that 100% of the fuel rods fail thereby releasing all of the available fission gas in the fuel rod gas gap to the cask cavity.

Calculation of the fission gas inventory is discussed in Section 4.2.1.1.

Page 4-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Source Terms Similar to normal transport conditions described in Section 4.2.1, the following equations from NUREG/CR-6487 [4] are used to determine the source term available for release.

Cvolatiles = {NAfBAVfv} / V Cgases = {NA fB AG fG } / V Cfines = {NA fB AF fF } / V Ccrud = {fC SC NR NA SAR } e-t / V Ctotal = Ccrud + Cvolatiles + Cgases + Cfines Table 4-1 shows the design basis radionuclide inventory corresponding to one fuel assembly for the containment evaluation. Table 4-2 shows the activity concentration from each of the sources available for release from inside the TN-40. The release fractions for the radionuclides are taken from NUREG/CR-6487. Under hypothetical accident conditions, the cladding of 100% of the fuel rods is assumed to fail (fB=1.0).

Containment of Radioactive Material The TN-40 is designed to meet the hypothetical accident requirements of 10 CFR 71.51. The A2 values are calculated using the methodology of 10 CFR 71.71 and NUREG/CR-6487. The A2 values are provided in Table 4-1 and Table 4-3.

Containment Criterion The allowable leak rates under hypothetical accident conditions are calculated using the methodology of NUREG/CR-6487 and previously presented in Section 4.2.1.3. The permissible leak rate under hypothetical accidents is 4.11E-02 cm3/sec (Table 4-4).

This value is converted to units of ref-cm3/ sec by first calculating the equivalent hole size. The equations of ANSI N14.5 (also see Section 4.2.1.3) are used:

Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec at Tu, Pu where:

Lu = 4.11E-02 cm3/sec Pu = 5.50 atm abs (conservative value)

Pd = 1.0 atm abs a = 0.5 cm

= 0.030 cP T = 531F (= 277C = 550 K) conservative value M = 4 g/mol (from ANSI N14.5, Table B.1)

Pa = 3.25 atm abs Page 4-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Substituting into the equations of ANSI N14.5:

Fc = (2.49x106 D4)/(0.5 0.030) = 1.66E+08 D4 Fm = {3.81x103 D3 (550/4)0.5} / {0.5 3.25} = 2.75E+04 D3 Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) 4.11E-02 = (Fc + Fm) (5.50 - 1.0) (3.25 / 5.50)

Fc + Fm = 1.55E-02 Solving the equations above for D, yields a hole diameter of 3.07E-03 cm.

This equivalent hole size, is then used to calculate the reference air rate at standard conditions. Assuming all upstream test conditions correspond to standard conditions:

Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec at Tu, Pu where:

Pu = 1.0 atm abs Pd = 0.01 atm abs D = 3.07E-03 cm a = 0.5 cm

= 0.0185 cP (from ANSI N14.5, Table B.1)

T = 298K M = 29.0 g/mol (from ANSI N14.5, Table B.1)

Pa = 0.505 atm abs Lu = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec Fc = {2.49x106 (3.07 x 10-3) 4}/(0.5 0.0185) = 2.40E-02 Fm = {3.81x103 (3.07 x 10-3)3 (298/29.0)0.5} / {0.5 0.505} = 1.40E-03 Lstd = (Fc + Fm)(Pu - Pd)(Pa/Pu) cm3/sec Lstd = (2.40E-02 + 1.40E-03)(1.0 - 0.01)(0.505 / 1.0)

Lstd = 1.27E-02 ref cm3/s Because the reference leak rate for normal conditions is lower than that for accident conditions, the leak test criterion developed in Section 4.2.1.3 demonstrates that the containment criteria for both normal and accident conditions are met.

The structural and thermal consequences of hypothetical accident loading conditions do not adversely affect the performance of the containment boundary structure or seals.

The impact limiters remain in place on the cask after the hypothetical accident as concluded in Appendix 2.10.8 for the 30 foot drop orientations. During the hypothetical accident the impact limiters provide insulation for the seals of the penetrations underneath them, including the lid seal, vent and drain ports, and the OP port.

Page 4-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Chapter 3 Table 3-3 lists the maximum temperature of the seals during a hypothetical thermal accident. Temperatures are shown for those areas protected by the insulating effect of the impact limiters, and other areas exposed directly to the accident temperatures environment. None of these temperatures exceeds the seal limit of 536F. The pressure inside the cask cavity also remains well below the design pressure of 100 psig as shown below (assuming 100% fuel rod failure).

PHAC = Pinitial + P100% rod failure Pinitial = (2.0 atm abs)(500 K / 449 K) = 2.23 atm abs P100% rod failure = (0.404 kgmole/cask 0.08314 bar-m3/kgmole K500 K) / (6.46 m3)

P100% rod failure = 2.6 bar = 2.57 atm abs PHAC = 2.23 + 2.57 = 4.80 atm abs = 70.6 psia = 55.9 psig The cavity gas mixture under accident conditions is presented in Table 4-6. The cavity gas mixture (assuming 100% fuel rod failure) consists of 60.5 % helium (from cask backfill operations and from rod pre-pressurization), 34.3% xenon, 3.4% krypton, 1.5%

iodine. This gas mixture is not explosive.

Leakage Rate Tests for Type B Packages Leakage tests performed on the TN-40 are based on those listed in Chapter 7 of ANSI N14.5 1997 [6]. A description of these tests is given below.

Design Leakage Rate Test: The containment boundary has been designed with appropriate materials and dimensions to provide containment for normal and accident conditions of transport. Analyses described within this SAR demonstrate the containment boundary design is acceptable for both normal and accident conditions of transport.

Fabrication Leakage Rate Test: As described in SAR Section 8.1.3, once during the fabrication of each packaging, a fabrication leakage rate test is performed with an acceptance criteria of 1x10-4 ref cm3/s. Typically, helium mass spectrometer leakage testing is performed with a sensitivity of 5x10-5 ref cm3/s or less.

Pre-shipment Leakage Rate Test: Once prior to shipment of the packaging, a leakage rate test is performed as described in SAR Section 7.1. Typically, a helium mass spectrometer test is performed with an acceptance criterion of 1x10-4 ref cm3/s and a sensitivity of 5x10-5 ref cm3/s or less. Since the metallic seals are not used for more than one transport, the pre-shipment testing also fulfills the requirements for the maintenance and periodic leakage rate tests.

Page 4-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Special Requirements Per the requirements of 10 CFR 71.63 [2], shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than 0.74 TBq (20 Ci) of plutonium, TN-40 transport packaging has plutonium in the solid form in the fuel rods of spent fuel assemblies and therefore meets this requirement.

References

1. High Performance Sealing Metal Seals Catalogue, Helicoflex Co., Boonton, NJ, ET 507E5930.

2. 10 CFR 71, Packaging and Transportation of Radioactive Materials.

3. ASME Boiler and Pressure Vessel Code, 1989 Code without Addenda.

4. NUREG/CR-6487, Containment Analysis for Type B Packages used to Transport Various Contents, Lawrence Livermore National Laboratory, 1996.

5. TN40 SAR, Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Rev. 10.

6. ANSI N14.5-1997, American National Standard for Radioactive Material - Leakage Tests on Packages for Shipment, February 1998.

7. NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, March 2000.

8. SCALE-4.4, Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL.

Page 4-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-1 Radionuclide Inventory and A2 Values Ai1 FA A2 Value FA / A2 Nuclide (Ci/assy) Fraction (Ci) (Ci-1)

Crud co60 4.54 1.0000 11.0 9.09E-02 Total - Crud 4.54 1.0000 11.0 1/Sum(FA/A2)

Volatiles cs1372 72300 0.7381 16 4.61E-02 cs134 555 0.0057 19 2.98E-04 sr90 25100 0.2562 8.1 3.16E-02 Total - Volatiles 97955 1.0000 12.8 1/Sum(FA/A2)

Gases i129 0.02 8.9E-06 Unlimited 0.00E+00 kr 81 4.66E-08 2.7E-11 1100 2.44E-14 kr 85 1630 0.9406 270 3.48E-03 rb 87 9.72E-06 5.6E-09 Unlimited 0.00E+00 h3 103 0.0594 1100 5.40E-05 Total - Gases 1733 1.0000 283 1/Sum(FA/A2)

Fines pu238 1380 0.0217 0.027 8.05E-01 pu239 137 0.0022 0.027 7.99E-02 pu240 217 0.0034 0.027 1.27E-01 pu241 31600 0.4975 1.6 3.11E-01 am241 1140 0.0179 0.027 6.65E-01 am243 14.1 0.0002 0.027 8.22E-03 cm243 6.5 0.0001 0.027 3.79E-03 cm244 1130 0.0178 0.054 3.29E-01 np239 14.1 0.0002 11 2.02E-05 eu154 959 0.0151 16 9.44E-04 eu155 150 0.0024 81 2.92E-05 pm147 1500 0.0236 54 4.37E-04 sm151 165 0.0026 270 9.62E-06 y 90 25100 0.3952 8.1 4.88E-02 Total - Fines 63512.7 1.0000 0.420 1/Sum(FA/A2)

Hypothetical Accident (Gases) i129 1.55E-02 1.5E-04 Unlimited 0.00E+00 kr 81 4.66E-08 4.5E-10 1100 4.11E-13 rb 87 9.72E-06 9.4E-08 Unlimited 0.00E+00 h3 1.03E+02 0.9998 1100 9.09E-04 Total - Gases 1.03E+02 1.0000 1100 1/Sum(FA/A2) 1 Values are based on a 14x14 WE STD fuel assembly (39,000 MWD/MTU burnup, 3.3 wt% U-235 initial bundle average enrichment, and 15 year cooled).

2 Ba137m contributes about 21% to the total design basis activity. Ba137m is a daughter of Cs137 with a half life of 2.6 min. In accordance with 10CFR71 Appendix A Note III, this radionuclide is evaluated with the parent nuclide.

Page 4-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-2 Activity Concentration by Source Activity Fraction available for Fraction of rods Concentration Source release from the fuel rod(1) that develop in TN-40 cask (fV / fG / fF / fC) cladding breach(1)

(Ci/cm3)(2,3)

Normal Transport Conditions Volatiles 2E-04 0.03 3.64E-06 Gases 0.3 0.03 9.65E-05 Fines 3E-05 0.03 3.54E-07 Crud(4) 0.15 not applicable 4.21E-06 Hypothetical Accident Conditions Volatiles 2E-04 1.0 1.21E-04 Gases 0.3 1.0 1.91E-04 Gases - Kr-0.3 1.0 3.03E-03 85 only Fines 3E-05 1.0 1.18E-05 Crud 1.0 not applicable 2.81E-05 1 Values taken from NUREG/CR-6487 [4].

2 40 assemblies per cask.

3 Cavity free volume is equal to 6.46E+06 cm3 4 Crud source is based on a surface area of 1.30E+03 cm2 / rod and an initial surface activity of 1.40E-04 Ci/cm2 at the time of discharge. At discharge, typically, fuel crud is composed of isotopes of cobalt, manganese, chromium and iron. After a 15 year cooling time, the only isotope of radiological significance is Co-60. A decay factor of 0.14 is included in the values listed above.

Page 4-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-3 Normal Conditions of Transport and Hypothetical Accident Conditions Effective A2 Values Fraction Effective A2 FA / A2 i Ci (Ci/cm3)

FA (Ci) (Ci-1)

Normal Conditions of Transport V,

3.64E-06 3.47E-02 12.8 2.71E-03 Volatiles G, Gases 9.65E-05 9.22E-01 283 3.26E-03 F, Fines 3.54E-07 3.38E-03 0.420 8.03E-03 C, Crud 4.21E-06 4.02E-02 11.0 3.66E-03 Total 1.05E-04 1.00E+00 56.6 1/Sum(FA/A2)

Hypothetical Accident Conditions V,

1.21E-04 3.44E-01 12.8 2.69E-02 Volatiles G, Gases 1.91E-04 5.43E-01 1100 4.93E-04 F, Fines 1.18E-05 3.35E-02 0.420 7.96E-02 C, Crud 2.81E-05 7.97E-02 11.0 7.25E-03 Total 3.52E-04 1.00E+00 8.8 1/Sum(FA/A2)

Page 4-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-4 Normal Condition of Transport and Hypothetical Accident Conditions Permissible Leakage Rates from the TN-40 Permissible Effective Allowable Allowable Concentration Leakage Permissible A2 Release Release Rate Ci Rate Std Leakage Rate Case (Ci) Rate (Ci/sec) (Ci/cm3) (cm3/sec) (ref cm3/sec)

NCT 56.6 A2 x 10-6 1.57E-08 1.05E-04 1.50E-04 1.0E-04 per hour HAC 8.8 A2 per week 1.45E-05 3.52E-04 4.11E-02 1.27E-02 Kr-85 270 10A2 per 4.46E-03 3.03E-03 1.48E+00 Note 1 (HAC) week Note (1): Hypothetical accident conditions without Kr-85 are bounding. This value is not calculated.

Page 4-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-5 Total Moles of Fission Gas for a WE 14x14 Std. Assembly WE 14x14 std - Design Basis Fuel for 45 GWD/MTU 3.8% U-235 15 yrs decay Containment SAS2H / M m/M Release M (molar ORIGEN-S (kgrelease / (kgmolerelease Fraction mass)

(g/assembly) assembly) / cask)

Actinides he4 9.030E-01 0.3 2.709E-04 4 2.709E-03 Total He: 2.709E-03 Fission h 3 1.210E-02 0.3 3.630E-06 3 4.840E-05 Products Total H: 4.840E-05 i127 2.370E+01 0.3 7.110E-03 127 2.239E-03 i129 1.000E+02 0.3 3.000E-02 129 9.302E-03 Total I: 1.154E-02 kr 80 3.160E-05 0.3 9.480E-09 80 4.740E-09 kr 82 4.170E-01 0.3 1.251E-04 82 6.102E-05 kr 83 2.100E+01 0.3 6.300E-03 83 3.036E-03 kr 84 6.470E+01 0.3 1.941E-02 84 9.243E-03 kr 85 4.760E+00 0.3 1.428E-03 85 6.720E-04 kr 86 1.010E+02 0.3 3.030E-02 86 1.409E-02 Total Kr: 2.711E-02 xe128 1.760E+00 0.3 5.280E-04 128 1.650E-04 xe129 1.350E-02 0.3 4.050E-06 129 1.256E-06 xe130 5.100E+00 0.3 1.530E-03 130 4.708E-04 xe131 2.130E+02 0.3 6.390E-02 131 1.951E-02 xe132 6.390E+02 0.3 1.917E-01 132 5.809E-02 xe134 8.460E+02 0.3 2.538E-01 134 7.576E-02 xe136 1.320E+03 0.3 3.960E-01 136 1.165E-01 Total Xe: 2.705E-01 Total: 3.119E-01 Page 4-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 4-6 Cask Gas Mixtures under Normal Conditions of Transport and Hypothetical Accident Conditions NCT HAC (3% rod failure) (100% rod failure)

(Kg-mole/cask) (Kg-mole/cask)

Fission Products I 3.46E-04 1.15E-02 Kr 8.13E-04 2.71E-02 Xe 8.11E-03 2.70E-01 He 8.13E-05 2.71E-03 Subtotal 9.35 E-03 3.09 E-03 Pre-pressurization He 2.75E-03 9.167E-02 Helium Backfill He 3.858E-01 3.858E-01 Total 3.979E-01 7.894E-01

% Mass of Gases Fission Products I 0.1% 1.5%

Kr 0.2% 3.4%

Xe 2.0% 34.3%

He - 0.3%

Pre-pressurization He 0.7% 11.6%

Helium Backfill He 97.0% 48.9%

Total 100.0% 100.0%

Page 4-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Notes to Figure 4-1:

1. Figure not to scale. Features exaggerated for clarity.

2. Phantom lines ( ) indicate containment boundary.

3. Containment boundary components are listed below:

1 Inner shell.

2 Lid outer plate, closure bolts and inner o-rings.

3 Shell flange.

4 Vent port cover plate, bolts and seals.

5 Drain port cover plate, bolts and seals (not shown).

Figure 4-1 TN-40 Containment Boundary Components Page 4-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

.260 (LID), .161 (VENT/DRAIN)

Figure 4-2 Lid, Vent Port And Drain Port Metal Seals Page 4-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 5 SHIELDING EVALUATION TABLE OF CONTENTS Discussion and Results........................................................................... 5-1 Source Specification ............................................................................... 5-3 Axial Source Distribution .............................................................. 5-4 Gamma Source ............................................................................ 5-5 Neutron Source ............................................................................ 5-6 Source Conversion Factors .......................................................... 5-6 Fuel Qualification ......................................................................... 5-7 Model Specification ............................................................................... 5-10 Description of Radial and Axial Shielding Configuration ............ 5-11 Shield Regional Densities .......................................................... 5-12 Shielding Evaluation ............................................................................. 5-12 Uncertainties and Conservatism in the Shielding Evaluation ................ 5-15 References ........................................................................................... 5-17 Input File Listing .................................................................................... 5-18 SAS2H/ORIGEN-S Input File ..................................................... 5-18 MCNP Neutron Model Input File ................................................ 5-19 MCNP Primary Gamma Input File .............................................. 5-38 Page 5-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 5-1 TN-40 Cask Shield Materials ................................................................ 5-58 Table 5-2 Summary of TN-40 Dose Rates ............................................................ 5-59 Table 5-3 PWR Fuel Assembly Design Characteristics ........................................ 5-60 Table 5-4 Westinghouse 14 X 14 STD Fuel Assembly Hardware Characteristics ...................................................................................... 5-61 Table 5-5 Material Compositions for Fuel Assembly Hardware Materials ............. 5-62 Table 5-6 Source Distribution ............................................................................... 5-63 Table 5-7 TPA Gamma Source ............................................................................. 5-63 Table 5-8 Fuel Qualification Table ........................................................................ 5-64 Table 5-9 Minimum Cooling (Years) Required to Meet Radiation and Decay Heat Limits ................................................................................. 5-65 Table 5-10 Dose Rates (mrem/hr) at 2 Meters from Side of 10 wide Transportation Platform Estimated with Response Function ................ 5-66 Table 5-11 Decay Heat Output (kW PER CASK) .................................................... 5-67 Table 5-12 Axial Source Term Peaking Summary .................................................. 5-68 Table 5-13 Fuel Assembly Materials Input For MCNP ............................................ 5-69 Table 5-14 Package Materials Input for MCNP ....................................................... 5-71 Table 5-15 Flux-to-Dose Rate Conversion Factors for Gamma .............................. 5-72 Table 5-16 Flux-Dose-Rate Conversion Factors for Neutron .................................. 5-73 Table 5-17 Average End Dose Rates as a Function of Railcar Length ................... 5-74 Table 5-18 Total Dose Rates along Side of the Cask at Normal Conditions-Design Basis ......................................................................................... 5-75 Table 5-19 Total Dose Rates along Side of the Cask at Normal Conditions-30 Year Cooled ..................................................................................... 5-76 Table 5-20 Response Functions Employed for Fuel Qualification .......................... 5-77 LIST OF FIGURES Figure 5-1 Cask Shielding Configuration................................................................ 5-78 Figure 5-2 Axial Burnup Profile for Design Basis Fuel ........................................... 5-79 Figure 5-3 Side View of TN-40 Transport MCNP Model ........................................ 5-80 Figure 5-4 Detail Views of TN-40 Transport MCNP Model ..................................... 5-81 Figure 5-5 Plan View of TN-40 Transport MCNP Model Basket Structure ............. 5-82 Figure 5-6 Details of Lattice Unit Cell and Rails/Outer SST ................................... 5-83 Page 5-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 5-7 DELETED ............................................................................................. 5-84 Figure 5-8 DELETED ............................................................................................. 5-84 Page 5-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 5.0 SHIELDING EVALUATION Discussion and Results Shielding for the TN-40 package is provided mainly by the cask body. The cask body is made up of the containment vessel, the gamma shielding and the lid. For the neutron shielding, a borated polyester resin compound surrounds the gamma shield shell radially. Additional shielding is provided by the steel outer shell surrounding the resin layer and by the steel and aluminum structure of the fuel basket.

For transport, wood filled impact limiters are installed on either end of the cask and provide additional shielding for the ends and some radial shielding for the areas at either end of the radial neutron shield. Figure 5-1 shows the configuration of the package shielding. Table 5-1 lists the compositions of the shielding materials.

The fuel assemblies acceptable for transport in the TN-40 are listed in Section 1.2.3.

Using the SAS2H/ORIGEN-S modules of SCALE [1], source terms are calculated. The bounding design basis fuel for dose rate has an initial enrichment of 2.35 wt% and a total maximum bundle-average burnup of 42,000 MWD/MTU with a 24.4 year decay time. Note that the criticality evaluation documented in Chapter 6 requires a minimum cooling time of 30 years. The evaluated decay heat is 21 kW/cask as opposed to the decay heat of 19 kW/cask (corresponding to a cooling time of 30 years).

The Westinghouse 14x14 standard fuel assembly contains the maximum heavy metal weight (Section 1.2.3) which results in bounding neutron and gamma source terms and is therefore identified as the most conservative fuel assembly. Section 5.2 describes the source specification and Section 5.4 describes the shielding analysis performed for the TN-40 cask. The shielding analysis models are described in Section 5.3.

Normal Conditions of Transport (NCT) are modeled with the neutron shielding and impact limiters intact. This shielding calculation is performed using the Monte Carlo computer code MCNP [5, 9]. Dose rates on the side, top and bottom of the TN-40 package are calculated for the various sources described in Section 5.2 and summed to give a total gamma and neutron dose rate.

Hypothetical Accident Conditions (HAC) assume that the neutron shield and the impact limiters are removed. This evaluation bounds the accident conditions since it is shown in Chapter 2 and Chapter 3 that the neutron shielding may be lost but the impact limiters remain on the cask during HAC. Shielding calculations for the HAC are also performed using MCNP.

Page 5-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The expected maximum dose rates (for NCT and HAC) from the TN-40 package are provided in Table 5-2. These dose rates are the design basis dose rates for the TN-40 package with a minimum cooling time of 30 years. Although this dose rate evaluation is performed using design basis fuel, fuel qualification evaluations were performed to determine that 15 year minimum cooled fuel is also acceptable for certain burnup and enrichment combinations. These evaluations were performed to determine the fuel assembly parameters of burnup, percent initial enrichment and cooling time that would result in decay heat and radiological sources that would meet the decay heat requirements (Chapter 3), source terms for containment (Chapter 4) and radiological sources that provide dose rates less than the current design basis fuel mentioned above and thus would be acceptable for transport (from a shielding standpoint) in the TN-40 package. Section 5.2 describes these evaluations in more detail.

The shielding calculations considered effects of tolerances. Since dose rates along the side of the transportation package are controlling, the cumulative effect of tolerances

(+.05/-.01 in. on 1.50 in. thick inner shell and +/-.12 in. on 8.00 in. thick gamma shell) of steel thicknesses and tolerances (+/-.12 in. on 4.50 in.) of resin on the side of the cask is considered. Only tolerances in thicknesses of the neutron shielding, cask inner and gamma shells affect the dose rates along the side of the cask. Note, dose rates presented in Table 5-2, Table 5-18, and Table 5-19 include the effect of the described geometrical tolerances.

The effect of material tolerances is considered only for the neutron shielding resin. Only hydrogen concentration is considered significant enough to affect the dose rates. The weight percent of hydrogen considered in the design basis shielding analysis represents the minimum guaranteed composition following the resin qualification. The average measured hydrogen weight percent in the TN-40 casks is 5.21 while that employed in the calculations is 5.05. The boron content has an effect on the secondary gamma dose rate. However, a concentration of greater than 0.75% ensures that this concentration is saturated and is sufficient to reduce the contribution of the secondary gamma component. Therefore, a material tolerance calculation with boron is not performed.

The effect of the tolerances on dose rates at various distances from the ends and at radial distances from the side greater than 2 meters is not significant.

The following items are also considered when using dose rates in Table 5-2, Table 5-18, and Table 5-19.

Design basis Westinghouse 14x14 Standard fuel assemblies with the bounding neutron and gamma source terms are utilized in the shielding evaluation.

The fuel qualification methodology calls for conservatively adjusting the enrichment/burnup and cooling time of the loaded fuel assemblies (Table 5-8).

Calculated dose rates are generally higher than measured dose rates, demonstrating the conservatisms in the shielding analysis methodology.

Page 5-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

The burnup-enrichment parameters of the design basis fuel assembly employed in the shielding evaluation are conservative compared to the burnup-enrichment distribution for the actual inventory of fuel assemblies as shown in Figure 6-17.

Source Specification There are five principal sources of radiation associated with transport of spent nuclear fuel that are of concern for radiation protection:

Primary gamma radiation from spent fuel,

Primary neutron radiation from spent fuel (both alpha-n reactions and spontaneous fission),

Gamma radiation from activated fuel structural materials and fuel inserts,

Capture gamma radiation produced by attenuation of neutrons by shielding material of the cask, and

Neutrons produced by sub-critical fission in fuel.

The TN-40 package is designed to transport Westinghouse 14 x 14 class PWR fuel assemblies. The fuel assemblies acceptable for transport in the TN-40 are described in Section 1.2.3. The various fuel assembly designs were separated according to fuel assembly array, the maximum metric tons of uranium, and the number of guide

/instrument tubes. These parameters are the significant contributors to the SAS2H/ORIGEN-S model. The largest uranium loading results in the largest source term at the design basis enrichment and burnup, thus the Westinghouse 14 x 14 standard is the bounding assembly type.

Table 5-3 provides characteristics of the design basis fuel assembly used in the source term analyses. The SAS2H/ORIGEN-S modules of the SCALE code are used to generate gamma and neutron source terms for the bounding Westinghouse 14 x 14 standard assembly. Source terms were generated for initial enrichments ranging from 2.00 wt% to 3.85 wt% U235 and the fuel is irradiated for a constant time of 400 effective full power days per cycle. Burnup values range from 17 GWD/MTU to 45 GWD/MTU using a specific radiation power between about 15 and 25 MW/assembly. A conservative operating cycle history is utilized with a 30 day down time between cycles.

Details of the analyses are given in Section 5.2.5.

The source terms are generated for the fuel assembly active fuel region, the plenum region, and the end fitting regions. The fuel assembly hardware materials and masses on a per assembly basis are listed in Table 5-4. Table 5-5 provides the material composition of fuel assembly hardware materials. Cobalt impurities in steel (800 ppm) and Inconel (4690 ppm) included in the SAS2H model are obtained from Table 5.1 of reference [2].

Page 5-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The masses for the materials in the top end fitting, the plenum, and the bottom fitting regions are multiplied by 0.1, 0.2 and 0.2, respectively [4]. These factors are used to correct for the spatial and spectral changes of the neutron flux outside of the active fuel zone. The material compositions of the fuel assembly hardware are included in the SAS2H/ORIGEN-S model on a per assembly basis. The data shown in reference [4]

indicates that the cobalt impurities in steel and Inconel are 1500 ppm and 1300 ppm respectively. Using cobalt impurity levels shown in Table 5-5 yields a total cobalt content per fuel assembly that is slightly greater than that calculated using the impurity levels from reference [4]. When the cobalt content is adjusted using the above described flux factors, the resulting cobalt content employed herein is higher by approximately 50% indicating that the source terms are calculated using a very conservative representation of total cobalt content.

Using the same approach, the total cobalt content employed in the plenum, top and bottom end fitting regions is lower by a factor of 1.2, 1.4, and 1.9, respectively. This indicates that the dose rates calculated for the top and bottom ends of the package could potentially be under-estimated. However, this potential non-conservatism is offset by the use of a 30-year cooling time which is sufficient to reduce the cobalt source term in the end fittings. Further, any contribution from the active fuel region (conservative by a factor of 1.7) to the top and bottom end dose rates will serve to reduce this potential non-conservatism due to the use of a significantly higher cobalt source.

Axial Source Distribution PWR plant operations data for over twenty 14 x 14 fuel assemblies with approximately 36 to 42 GWD/MTU burnup are averaged into a typical profile, shown as maximum profile in Figure 5-2. Also shown in Figure 5-2 is the axial profile from Reference [3] for 38-42 GWD/MTU burnup fuel. The third profile shown in Figure 5-2 is a bounding profile and used in this analysis. The bounding profile is also applicable for fuel with blankets at the ends of the active zone. The use of a bounding profile ensures that the most penalizing profile is utilized that includes and accounts for both blanketed and un-blanketed fuel. The bounding profile is derived from an evaluation that includes fuel assemblies with blankets. Enrichment used in the shielding analysis is minimum assembly average and accounts for blankets. Burnup used in shielding analysis is maximum assembly average and accounts for blankets.

Using a burnup profile accounting for the presence of blankets with low enriched or natural uranium increases dose rates on the side of the package.

The conservative axial profile containing 12 axial zones is utilized in the shielding evaluation. The top and bottom 17% of the assembly are divided into two zones each and the middle 66% are divided into 8 approximately equal zones. The peaking factors range from 0.700 at the bottom and top, to a maximum of 1.16 just below the middle.

The gamma source is directly proportional to the burnup and the neutron source is proportional to the fourth power of the burnup. This data is presented in Table 5-12.

Page 5-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Gamma Source The gamma source terms for the design basis spent fuel assembly is provided in Table 5-6. Table 5-6 presents the source terms for a Westinghouse 14 x 14 standard assembly with an initial enrichment of 2.35 wt%, maximum average burnup of 42,000 MWD/MTU, 24.4 year decay with a 13 year cooled TPA insert.

The gamma source spectra are presented in the 18-group structure consistent with the SCALE 27n-18 cross section library. The conversion of the source spectra from the default ORIGEN-S energy grouping to the SCALE 27n-18 energy grouping is performed directly through the ORIGEN-S code. The SAS2H/ORIGEN-S input file for this fuel assembly is provided in Section 5.6.

The gamma source for the fuel assembly hardware is primarily from the activation of cobalt. This activation contributes primarily to SCALE Energy Groups 36 and 37. The gamma source for the plenum region, the top fitting region and the bottom fitting region is provided in Table 5-6.

The spent fuel assemblies may contain irradiated fuel inserts (BPRA, TPA) which also provide a gamma source which is primarily from activated cobalt. The gamma source from a TPA corresponding to maximum host assembly burnup of 125,000 MWD/MTU and cooled for 13 years is shown in Table 5-7. This gamma source is added to the irradiated fuel gamma source in the plenum and top end fitting regions.

An axial burnup profile has been developed as discussed in Section 5.2.1 above. Table 5-12 provides design axial gamma peaking factors that were utilized in the MCNP shielding model.

Page 5-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Neutron Source Table 5-6 provides the total neutron source for the design basis fuel assembly under the irradiation/decay history described above in 5.2. The neutron source is comprised mainly of Cm-244 and for the MCNP analyses, the default Cm-244 energy spectrum was utilized.

The neutron source is not linearly dependent with burnup, and therefore calculations were performed to determine the axial neutron source distribution (Section 5.2.1). The axial neutron peaking factors are shown in Table 5-12.

Source Conversion Factors The following equation defines how the absolute tallies are calculated:

1000 mrem D = T S C BU 1 rem

Where, D is the absolute dose rate in mrem/hr, T is the MCNP tally result in rem/hr per particle/sec, S is the source strength in source particle/sec, and CBU is the axial burnup normalization constant.

In the above relationship the constants multiplied against the tally result, T, define the tally multiplier. Therefore, the tally conversion factor, CT, is defined below.

1000 mrem CT = S C BU 1 rem The source strength must be scaled appropriately by the axial burnup normalization constant. For the axial profile the axial burnup normalization constant is 1.049 (Table 5-12). This corresponds to the CBU for the gamma source. Using the fourth power approximation the neutron CBU is 1.367 (Table 5-12).

Neutron Source Term CT = [2.630E+08 neutron/sec/assy x 40 assy] x 1.367 x 1000 mrem/rem

= 1.438E+13 neutron-mrem/rem-sec Gamma Source Term CT = [1.787E+15 gamma/sec/assy x 40 assy] x 1.049 x 1000 mrem/rem

= 7.50E+19 gamma-mrem/rem-sec Page 5-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Fuel Qualification As stated previously, an evaluation was performed to determine the fuel assembly parameters of burnup, percent initial enrichment and cooling time that would result in dose rates and decay heats less than the design basis fuel mentioned above and thus would be acceptable for transport in the TN-40 cask.

These analyses were carried out using the SAS2H depletion module from the SCALE computer software and MCNP. For all SAS2H calculations the latest SCALE 44 group ENDF/B-V (44groupndf5) library was used. MCNP calculations used the default cross section libraries.

An MCNP model was utilized to calculate a response function at 2 meters from the transportation vehicle. Segmented surface tallies at 2 meters radial distance from the side of the transportation package (conservatively, 2m from an assumed 10 ft wide railcar) obtained from the MCNP analysis are essentially a source to dose conversion factor as a function of energy, (for gamma). These conversion factors are multiplied by an adjustment factor (identical to the axial burnup normalization constant, CBU, described in Section 5.2.4) to account for the axial burnup profile of the fuel to obtain the response functions that are used for the fuel qualification evaluation. The axial burnup profile utilized in the fuel qualification analysis is the same as the one utilized in the design basis dose rate shielding calculation. For neutrons, since a bounding energy spectrum is used, the response function calculated is just a total source to dose factor.

More than 200 SAS2H analyses were performed to determine gamma, neutron and thermal source terms as a function of burnup, enrichment and cool time. The gamma source was obtained as a function of energy. Note however that even though the source energies range between 0.05 and 10 MeV, the primary gamma radiation dose rate is essentially contributed (> 99%) by the source energy groups within the 0.4 to 4.0 MeV range. Response function entries were generated for each of the most contributing to the primary gamma radiation dose rate energy group. Two meter radial primary gamma dose rates were estimated/calculated by multiplying those entries and the appropriate SAS2H gamma energy source and performing a summation over the energy groups.

Since a bounding energy spectrum was used for the neutrons, the neutron and secondary gamma dose rates were calculated by multiplying the neutron response function and the total SAS2H neutron source and the secondary gamma response function and the neutron source. The estimated total 2 meter dose rate at approximately 20 cm above the middle of the active fuel was determined by combining the gamma and neutron dose rates. Two meters radial distance is considered from the side of the 10 wide transportation platform, which is 2 feet less than the width of the impact limiters. Provided the decay heat is less than 525 watts, the estimated dose rate must be less than 9.8 mrem/hr for the fuel to be qualified for transport. The calculated response functions are also conservatively increased by 5% to estimate the dose rates and are shown in Table 5-20.

Page 5-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The response function methodology employed to determine acceptable combinations of burnup, enrichment and cooling time (BECT) parameters for transport is described below:

MCNP models are developed - one each for gamma and neutron that represent the TN-40 transport package fully in 3D. The objective is to determine the dose rate contribution of radiation sources (both gamma and neutron) at 2 m from the radial surface of the package.

The axial ends of the cask and the resin cutout regions around the trunnions are not modeled exactly. However, these modeling considerations do not affect the response function results at the area of interest. Therefore, these models are considered essentially the same.

The fuel assembly is modeled as a homogenized cuboid and is divided into four axial regions - representing the bottom nozzle, in-core (or active fuel), the plenum and the top nozzle regions.

Because the intensity of the gamma sources in the in-core region is higher than that of the other three regions by approximately a factor of 1000, the axial source distribution in the MCNP models (for gamma) is set to this ratio. Note that this ratio is only employed to distribute the combined gamma source within the various regions of the fuel assembly. The spectral distribution is input separately and is not affected by this simplification. This inherently assumes that the resulting response functions are accurate and that the source distribution intensities of these regions are relatively representative. This assumption is justified for cooling times greater than or equal to 10 years due to the decay of Co-60 and the reduction in the source term intensity in the fuel hardware regions.

In addition, the response functions are calculated at a radial distance of 2 m from the mid-plane axial location of the active fuel where the contribution of the in-core source region dominates that of the hardware regions. Therefore, the expected changes in the relative source intensities of the hardware regions do not affect the dose rates significantly.

For gamma, the response functions (dose rate per source particle) for nine energy groups are calculated using a single MCNP model. These response functions also include the contribution from the end fitting regions. Further, these response functions are increased by 5% to account for additional uncertainty. These response functions are shown in Table 5-20.

For neutrons and secondary gamma, a single response function is calculated since the MCNP built-in spectrum was utilized in these calculations. These response functions are also increased by 5% to account for additional uncertainty and are also shown in Table 5-20.

Once the response functions are calculated, source terms are generated for the various burnup and enrichment combinations to determine the required cooling times for transport.

Page 5-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

SAS2H/ORIGEN-S is employed to determine the source terms employed in the fuel qualification calculations. Since the gamma response functions are calculated for a combined fuel assembly source, the SAS2H/ORIGEN-S calculations are also performed to determine source terms that include the contribution from all the four regions. Note that the neutron sources are not applicable to fuel hardware regions.

Numerous BECT combinations are utilized to determine the decay heat and radiation source terms. The dose rates from response functions are then estimated by adjusting the cooling times to ensure that the resulting dose rates are less than 9.8 mrem/hour. Cooling times are also adjusted such that the resulting decay heat values are below 525 watts per fuel assembly. The results of these calculations are shown in Table 5-9 where the minimum required cooling times are calculated as a function of burnup and enrichment.

The calculated dose rate and decay heat along with the cooling time are then utilized according to the steps above to determine the bounding radiological source term. The final design basis radiological source term was generated by adding the TPA source term to the fuel/hardware source term because the BPRA source term is already included. Note that the TPA source term was not explicitly included in the cooling time calculations. The cooling times calculated are reduced to a simplified look up table as a function of spent fuel parameters to summarize the loading parameters for the TN-40 transport package and are shown in Table 5-8.

Table 5-9 shows the results of the evaluation which define the spent fuel assembly cooling times to meet radiological and decay heat limits necessary for burnups ranging from 17 GWD to 45 GWD and enrichments between 2.0 wt% and 3.85 wt%. The TN-40 package containing fuel assemblies with parameters defined in this table will meet the dose rate and thermal criteria for transport. Table 5-10 shows the estimated dose rates and Table 5-11 shows the calculated decay heat corresponding to the cooling times shown in Table 5-9. All assemblies producing a decay heat of less than 21 kW per package or 525 watts per assembly are radiation (dose rate) limited. A fuel qualification table (FQT) for loading purposes based on this evaluation is provided in Table 1-2 (also shown in Table 5-8). The FQT is generated by conservatively rounding the cooling times shown in Table 5-9 up to the nearest value greater than 15 years.

The response function calculated dose rates shown in Table 5-10 are categorized into two sets: radiation limited and thermal limited. The dose rates for the radiation limited BE combinations are indicated by using a value of 10 mrem/hour while those for the thermal limited BE combinations are shown by using a value of less than 10 mrem/hour.

The FQT is developed such that any BECT combination shown in Table 5-9 that results in an estimated dose rate of 10 mrem/hour (rounded up) can be used in the shielding calculations. However, the bounding parameters are burnup of 42 GWD/MTU at the enrichment of 2.35 wt% U-235 and a cooling time of 24.4 years to calculate the design basis source terms.

Page 5-9

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 In order to ensure that the burnup and enrichment combinations employed previously also remain bounding for a cooling time of 30 years, a simple calculation is performed as described below. Based on the fuel qualification shown in Table 5-8, it is sufficient to demonstrate that the source terms (and the resulting dose rates) calculated using an enrichment of 2.35 wt% U-235 and a burnup of 42 GWD/MTU (design basis) are greater than those calculated for the following two combinations:

1) enrichment of 3.00 wt% U-235 and a burnup of 43 GWD/MTU (combination 1)

2) enrichment of 3.25 wt% U-235 and a burnup of 45 GWD/MTU (combination 2)

All other combinations of burnup and enrichment are bounded by the above combinations since a uniform cooling time of 30 years is employed.

The neutron source terms for combination 1 are lower than the design basis by approximately a factor of 0.77 while the total primary gamma source terms are higher by approximately a factor of 1.05. The primary gamma dose rate estimated using the response function shown in Table 5-20 for combination 1 is slightly lower than that for the design basis.

The neutron source terms for combination 2 are lower than the design basis by approximately a factor of 0.81 while the total primary gamma source terms are higher by approximately a factor of 1.1. The primary gamma dose rate estimated using the response function shown in Table 5-20 for combination 2 is comparable (within 0.5%) to that for the design basis.

As discussed in Section 5.4, the neutron source term contribution to the total dose rate at the side is greater than 70%. Since the neutron source terms for the design basis is greater than that for these two configurations, it can be concluded that the design basis combination will result in bounding source terms for a cooling time of 30 years.

In summary, the source terms from the design basis combination are bounding for all fuel assemblies with acceptable burn-up enrichment combinations with cooling times greater than or equal to 30 years.

Model Specification The Monte Carlo computer code MCNP [5] is used for calculating the gamma and neutron doses in this analysis. A more advanced version of the code MCNP [9] is also employed to determine the dose rates for the models that include tolerances as described in Section 5.4.

Page 5-10

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Description of Radial and Axial Shielding Configuration Two base models were constructed. The first model corresponds to the neutron transport problem and the second is the gamma. Variance reduction was accomplished by means of importance zoning followed by weight windows. The importance function was created to balance the particles (per volume) throughout the problem geometry.

The process used to do this was an iterative approach starting with basic attenuation factors for the shielding materials. The neutron importance function developed was also applied to the secondary gammas.

The test importance functions were then run in conjunction with the weight window generator. The weight windows calculated were inserted into the final MCNP runs.

Weight windows were only used in the gamma cases.

The models were used to calculate both the axial and radial dose rates. The impact limiters and radial neutron shielding were removed for the HAC evaluation.

Sections 5.3.1.1 and 5.3.1.2 describe the shielding model (for the vicinity immediately around the cask) developed for the TN-40 under NCT and HAC.

5.3.1.1 NCT Radial and Axial Shielding Configuration One shielding configuration is used for the TN-40 NCT design. The model is a complete three dimensional simulation of the TN-40 transportation package. The 72 inch diameter interior cavity of the cask is modeled with a discrete representation of the basket and fuel structure. Each fuel assembly is divided into four axial zones. The bottom zone represents the lower end fittings, the middle zone the active fuel region and the upper zones represent the plenum and upper end fittings, respectively. The axial locations of the plenum and the end fittings of the fuel assembly are similar to those provided in Reference [8]. The modeled active fuel length is 144 inches and the plenum length is 7.14 inches. The modeled bottom end fitting and top end fitting lengths are 3.08 inches and 3.5 inches, respectively. The fuel, end fittings and plenum are homogenized within the each assembly envelope and the axial length of their respective zones.

The basket structure is modeled as a 0.755 inch thick grid of aluminum and steel panels. The periphery of the basket is modeled by several peripheral basket universes to best represent the geometry. The TN-40 package model is illustrated in Figure 5-3 through Figure 5-6.

The impact limiters are modeled as wood surrounded by a 0.25 in. thick steel shell. The interior steel gussets are conservatively neglected. The wood is modeled mostly as redwood except two areas, as shown in drawings 10421-71-41 and -42 (Appendix 1.4.1), which are modeled as balsa. An aluminum spacer utilized under the top impact limiter is included in the model.

Page 5-11

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The TN-40 cask body is modeled using the appropriate materials and dimensions shown in the drawings in Appendix 1.4.1.

Voids are neglected within the fuel assembly. The voids within the cask cavity are modeled.

5.3.1.2 HAC Radial and Axial Shielding Configuration under Hypothetical Accident Conditions For the HAC evaluation, it is conservatively assumed that the top and bottom impact limiters are destroyed and are no longer attached to the cask. Also, the neutron shield is assumed to be removed. The remaining model utilizes the same regional densities and shield thickness as the model for NCT.

Shield Regional Densities For the MCNP model, four source areas, shown in Figure 5-3, are utilized: fuel zone, plenum, top and bottom end fitting. The sources are uniformly homogenized over the cross section and the appropriate zone length. The fuel basket is discretely represented by the stainless steel and aluminum plates that bound the fuel assemblies.

The radial resin and aluminum boxes are homogenized into a single composition based on the mass of each component. Measured dose rates around the TN-24P [7], the TN-40, and the TN-32 casks have shown no streaming effects because of the aluminum boxes. This is because the neutrons will not generally travel in a direct path, but scatter, such that the majority of the neutrons will not be able to travel through the aluminum box wall for the full 4.5 inches of resin box thickness. The materials input for the MCNP model is listed in Table 5-13 and Table 5-14.

Shielding Evaluation Dose rates around the TN-40 package are determined by choosing the most conservative source and using it within a three dimensional MCNP model. Dose rates were estimated both axially and radially outside the cask. Several tallies were used to accomplish this. All tallies are either surface (F2) or volume flux (F4) tallies and are converted into dose rates using energy dependent dose conversion factors listed in Table 5-15 and Table 5-16 [6]. The tallies are further scaled by the source conversion factors described in Section 5.2.4. The tally locations were chosen to best evaluate the external dose rate requirements of 10 CFR 71.47(b). Tally segmentation was used to analyze the surface dose rates for any spatial peaking. This was done both axially and radially. The segments were typically 20 to 25 cm in dimension. Note that dose rates were calculated at several locations from the ends of the cask to assess hypothetical occupational exposure using surface average tallies as a function of railcar length.

Also, local tallies were implemented to calculate the dose rates around the trunnions and above the neutron shield. The MCNP shielding evaluation accounts for subcritical neutron multiplication. An initial enrichment of 3.0% wt U-235 was utilized in the MCNP input file to provide a conservative subcritical multiplication source.

Page 5-12

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 For the dose calculation around the TN-40, the source is divided into four separate regions: fuel, plenum, top end fitting, and bottom end fitting. The model is utilized in two separate computer runs consisting of contributions from the following sources:

Primary gamma radiation from the active fuel and from activated hardware within the top end fitting, plenum region and bottom end fitting (axial and radial directions).

Neutron radiation from the active fuel region and secondary gamma radiation from neutron interactions.

The sources in the active fuel region (gamma and neutron) are modeled as uniform radially but vary axially. The sources in the structural hardware regions (plenum, top end fitting, and bottom end fitting) are modeled as uniform both radially and axially. The results from the individual runs are summed to provide the total gamma, neutron and total dose for the package.

The statistical uncertainties are generally less than 5% for the majority of tallies except for local tally bins and the accident results. For the accident the neutron end dose rates have the highest relative error around 10%. The statistical uncertainties associated with the neutron dose rates on the top and bottom impact limiter surface are high, but since they contribute less than 1% (less than 0.1 mrem/hr) to the total dose this is acceptable.

The terminology for the dose locations is as follows. On the side of the cask results are reported on the surface of the cask (contact), at vertical planes extending up from a 10 feet wide vehicle (vertical planes), at the diameter of the impact limiters to represent the top and bottom of the package (top/bottom), 1 meter from the steel cask body (1 meter accident) and 2 meters from the vertical planes.

The results indicate peaking near the top and bottom of the cask and streaming in the upper trunnion/above the neutron shield regions. These results are expected due to the reduced shielding in these areas. It was determined that the normal conditions peak external surface dose rate of 60 mrem/hr occurs just above the neutron shield. This is approximately a factor of 1.8 times higher than the average on the cask surface. The localized peaking at the top of the cask is due to the absence of the neutron shield at the top. Neutron streaming was observed through the trunnion itself. However, the total dose rates just outside the trunnion were nearly the same as those averaged around the entire circumference of the cask.

Table 5-2 presents the maximum calculated dose at contact, at the vehicles outer edge (assumed 10 ft wide vehicle), and at 2 m from the vehicles outer edge. The calculated total dose rates at the various locations around the package are presented in Table 5-2, Table 5-18, and Table 5-19.

For the HAC, Table 5-2 also presents the maximum calculated doses at 1 m from the cask body.

The dose rates for an individual at the end of the rail car are presented in Table 5-17.

These results are presented as a function of the length of the rail car.

Page 5-13

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 On average, the dose rates are dominated by the neutron source term. The results indicate that typically the total dose rates are comprised of 25% to 30% primary gamma, 15% (n,) and 55% to 60% neutron. However, the primary gamma source produces the majority of the dose rate at the ends of the package; the average contribution from primary gamma is in the range of 80% to 85%. This is a direct result of the neutron shielding from the wood in the impact limiters. As expected, the accident dose rates are produced mostly from the neutron (94%) source due to the loss of the neutron shielding material and the impact limiter.

Typical average (beyond 130 cm above and below the active fuel midplane) contact dose rates on the side of the cask are approximately 36 mrem/hr (~55% neutron). At 2 meters from the side of the vehicle surface the dose rate is approximately 7.8 mrem/hr which is comprised of 4.3 mrem/hr neutron, 0.8 mrem/hr (n,) and 2.7 mrem/hr gamma.

On the ends, the total contact dose rates are less than 7 mrem/hr with less than a 0.1 mrem/hr contribution from neutrons. All these dose rates are at the lower tolerance limits of the shielding materials thickness on side of the cask and are based on source terms calculated with a cooling time of 30 years. Addressed tolerances are specified at the end of Section 5.1.

Axial distribution of the total dose rate at various radial distances from the side of the transportation package is presented in Table 5-18 and Table 5-19. Note that the neutron shield extends from -187 cm to +205 cm axial range in the MCNP calculational model.

The table shows that there is a dose rate increase from +190 cm to +230 cm axial coordinate range when considering dose rates at radial distances not exceeding the radius of impact limiters. The dose rate at that location is larger than at the middle of the cask because of less steel shielding due to the flat area near the trunnions and the absence of neutron shielding. Further, the MCNP calculational model for the top and plenum regions includes the gamma sources from BPRAs and TPAs. The bottom trunnions and the casks side at axial coordinates less than -187 cm are encompassed by the bottom impact limiter and there are no BPRA/TPA sources in the bottom region (the ratio of top/bottom gamma source strength is roughly a factor of 1.2). Therefore, the increase in the dose rates near the bottom trunnions is lower.

Because of the lack of shielding (see Figure 5-3) and radiological source concentration near the top trunnions, dose rates were examined at axial coordinates around the trunnions. Specifically, dose rates near the top trunnions, on the flat part around the top trunnions as pointed out with the P1 callout on a sketch of Figure 5-4, are evaluated.

The contact dose rate at this point is 46.8 mrem/hr (32.2 mrem/hr neutron, 3.9 mrem/hr (n,g) and 10.7 mrem/hr gamma). At 2 meters radial distance measured from the side of impact limiters, the total dose rate is 6.75 mrem/hr.

Page 5-14

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The dose rates shown in Table 5-18 and Table 5-19 correspond to 24.4-year cooled and 30-year cooled fuel, respectively. Two sets of 2 m dose rates are shown in these tables, based on a vehicle width of 10 ft and the edge of the impact limiters (12 ft wide),

respectively. The results shown in Table 5-19 with the 30-year cooled fuel demonstrate that the dose rates (maximum 7.79 mrem/hr) are below the limits (10 mrem/hr) with sufficient margin for a 10 ft wide vehicle. As shown in Section 5.2.5, the source term for the 30-year cooled fuel is determined using the bounding combination of enrichment and burnup. This ensures that the dose rates shown in Table 5-19 are design basis dose rates for 30-year cooled fuel.

The dose rate analysis was performed using MCNP [5, 9]. Selected inputs for MCNP are included in Section 5.7.

Uncertainties and Conservatism in the Shielding Evaluation The shielding evaluation described in Section 5.4 is based on a conservative representation of the geometry, material and source description. This section provides a description of the uncertainties associated with the shielding evaluation and demonstrates that the evaluation sufficiently covers these uncertainties.

Uncertainties can be due to 1) methodology - directly as a result of employing a computer program and interpretation of results and 2) modeling - use of geometrical, material and source representations. Since the calculated dose rates are based on results from source term and shielding calculations, this discussion extends to uncertainties from both the source term and shielding calculations.

The SAS2H/ORIGEN-S modules of the SCALE computer code package [1] were employed to determine the design basis source terms. The SCALE package and the 44 Group ENDF-B(V) cross section library has been extensively benchmarked to determine the uncertainties associated with the methodology. The SAS2H/ORIGEN-S prediction of the concentrations of the principal radioactive isotopes for both gamma and neutron source terms has been determined to be within 10% of measured data implying that the methodology uncertainty associated with SAS2H/ORIGEN-S is acceptably low. This indicates that the SAS2H code predictions are generally accurate and that the resulting source terms are appropriate, if not conservative. Further, comparison of surface dose rates from numerous measurements for various TN casks also confirm the applicability of the source term calculation methodology.

The calculational uncertainty in the models is more than compensated for by the use of conservative modeling parameters. The main source of conservatism is in the specification of fuel assembly hardware materials (including cobalt content) and irradiation history.

Page 5-15

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The fuel qualification methodology described in Section 5.2.5 also includes conservatisms in the employment of response functions. The selection of the bounding BECT parameters for calculating source terms among the various candidate BECT combinations (since they all result in approximately the same dose rate at 2 m) has an uncertainty that is less than 2% since the estimated dose rates are within 0.1 mrem/hr of each other. This is more than compensated for by the use of a scaling factor of 1.05 in the response function calculations.

The use of a Monte Carlo method for shielding calculations will result in a methodology uncertainty that is based on the standard deviation associated with the results. The dose rates from the calculations have associated standard deviations within 2%. In general, the MCNP methodology uncertainty is within 5% and is directly related to the acceptance criteria of the results. The modeling uncertainty is accounted for by employing the minimum geometrical tolerances for major components important to shielding. Additional conservatism in the use of the minimum guaranteed hydrogen content in the neutron resin ensures that the uncertainty of the resulting dose rates is minimized.

In general, the uncertainty associated with the MCNP methodology is more than compensated for with the conservatisms in the shielding models.

In summary, the methodology uncertainty associated with the source term and shielding calculations is within 10% and is directly related to the acceptance criteria (code benchmarking) of these methods - SAS2H/ORIGEN-S and MCNP. The source term and shielding analyses have been performed to ensure that the uncertainty associated with modeling (geometry, materials and source specification) is bounded by the conservatisms discussed above.

In addition, the dose rates shown in Table 5-2, Table 5-18 and Table 5-19 include the effect of geometrical and material tolerances. Most importantly, an analysis of the expected inventory in Chapter 6 indicates that the spent fuel parameters (burnup and enrichment) employed in the source term calculations are conservative (such a fuel assembly does not exist). Finally, the dose rates calculated with a cooling time of 30 years (required by the criticality analysis) ensure that the calculated dose rates are well below the applicable regulatory limits (Table 5-19).

Page 5-16

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. SCALE 4.4, Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL.

2. S.B. Ludwig and J.P. Renier, Standard and Extended-Burn-up PWR and BWR Reactor Models for the ORIGEN2 Computer Code, ORNL/TM-11018, Oak Ridge National Laboratory, December, 1989.

3. NUREG/CR-6801 (ORNL/TM-2001/273), March 2003, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses.

4. Luksic, Spent Fuel Assembly Hardware: Characterization and 10 CFR 61 Classification for Waste Disposal, PNL-6906, UC-85, June 1989.

5. MCNP4C2, Monte Carlo N-Particle Transport Code System, Los Alamos National Laboratory, CCC 701, RSIC, June 2001.

6. American National Standard for Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants, ANSI/ANS-6.6.1-1977, American Nuclear Society, Illinois, 1977.

7. EPRI-NP-5128, The TN-24P PWR Spent-Fuel Storage Cask: Testing and Analyses, prepared by Pacific Northwest Laboratory, Virginia Power Company and EG&G Idaho National Engineering Laboratory, April 1987.

8. DOE/ET/47912-3 Vol. III, Domestic Light Water Reactor Fuel Design Evolution, Prepared for U. S. Department of Energy Savannah River Operations Office, September 1981.

9. MCNP/MCNPX - Monte Carlo N-Particle Transport Code System Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries, CCC-730, Oak Ridge National Laboratory, RSICC Computer Code Collection, January 2006.

Page 5-17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Input File Listing SAS2H/ORIGEN-S Input File

=sas2h parm='halt02,skipshipdata' NSP W STD (14x14) 410 kgU 2.35 w/o 42 GWD/MTU 44groupndf5 latticecell

' =========================================================================

' FUEL MATERIAL SPECIFICATION 1=FUEL ALWAYS, 2=CLAD ALWAYS, 3=MOD ALWAYS

' =========================================================================

' %TD T (K) ... Uranium wt%

uo2 1 0.9555 840 92235 2.35 92238 97.6183 92234 0.02092 92236 0.01081 END zirc4 2 1.0 620 end h2o 3 den=0.733 1.0 558 end boron 3 den=0.733 600-6 558 end

' =========================================================================

' CASK MATERIAL COMPOSITIONS, START AT 6

' =========================================================================

' homogenized basket/cavity

' VF=partial dens/homo dens of fuel zr 6 den=0.349 end fe 6 den=0.248 end al 6 den=0.261 end

' primary cask shielding fe 7 den=7.85 end

' neutron shielding resin and aluminum al 8 den=0.482 end c 8 den=0.499 end o 8 den=0.592 end h 8 den=0.07171 end b 8 den=0.01491 1.0 293 5010 19.7 5011 80.3 end end comp

' =========================================================================

' FUEL PIN PARAMETERS

' =========================================================================

' pitch fuelod mfuel mmod cladod mclad cladid mgap squarepitch 1.41224 0.929386 1 3 1.07188 2 0.948436 0 end

' more data szf=1.2 end

' =========================================================================

' FUEL ASSEMBLY PARAMETERS

' =========================================================================

npin/assm=179 fuelength=365.76 ncycles=2 nlib/cyc=1 printlevel=5 lightel=35 inplevel=1 numinstr=1 ortube=0.53594 srtube=0.474218 end

' =========================================================================

' FUEL IRRADIATION in MW/BASIS and DAYS

' =========================================================================

power=21.5250 burn=400 down=30 end power=21.5250 burn=400 down=3652 end Page 5-18

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

' =========================================================================

' LIGHT ELEMENT SECTION (KG/BASIS)

' =========================================================================

H 1.08E-3 Li 4.10E-4 B 4.38E-4 C 5.50E-2 N 3.40E-2 O 5.52E+1 F 4.39E-3 Na 6.15E-3 Mg 8.20E-4 Al 4.10E-2 Si 9.30E-2 P 1.78E-2 S 5.61E-3 Cl 2.17E-3 Ca 8.20E-4 Ti 4.50E-2 V 2.90E-3 Cr 2.59E+0 Mn 1.68E-1 Fe 6.49E+0 Co 3.26E-2 Ni 3.49E+0 Cu 7.44E-3 Zn 1.65E-2 Zr 8.16E+1 Nb 2.98E-1 Mo 1.65E-1 Ag 4.10E-5 Cd 1.03E-2 In 8.20E-4 Sn 1.34E+0 Gd 1.03E-3 Hf 6.50E-3 W 2.49E-3 Pb 4.10E-4

' =========================================================================

' RADIAL GEOMETRY OF CASK

' =========================================================================

end end

' =========================================================================

' END CARD TO TERMINATE SAS RUN

' =========================================================================

end

=ORIGENS 0$$ A8 26 A11 71 E 1$$ 1 1T decay 24.0 24.4 26.0 28.0 30.0 32.0 35.0 40.0 45.0 50.0 YEARS 3$$ 21 A3 1 A30 18 A33 18 T T

56$$ 0 10 A13 -1 5 E T 60** 24.0 24.4 26.0 28.0 30.0 32.0 35.0 40.0 45.0 50.0 61** F1 65$$ A25 1 E 65$$ A10 1 A31 1 A52 1 E 81$$ 2 0 26 1 E 82$$ F2 83** 1.00E+07 8.00E+06 6.50E+06 5.00E+06 4.00E+06 3.00E+06 2.50E+06 2.00E+06 1.66E+06 1.33E+06 1.00E+06 8.00E+05 6.00E+05 4.00E+05 3.00E+05 2.00E+05 1.00E+05 5.00E+04 1.00E+04 T 24 YEAR COOLING WE STD 14X14 410KGU 24.4 YEAR COOLING WE STD 14X14 410KGU 26 YEAR COOLING WE STD 14X14 410KGU 28 YEAR COOLING WE STD 14X14 410KGU 30 YEAR COOLING WE STD 14X14 410KGU 32 YEAR COOLING WE STD 14X14 410KGU 35 YEAR COOLING WE STD 14X14 410KGU 40 YEAR COOLING WE STD 14X14 410KGU 45 YEAR COOLING WE STD 14X14 410KGU 50 YEAR COOLING WE STD 14X14 410KGU 56$$ F0 6T End Page 5-19

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 MCNP Neutron Model Input File TN-40 TRAN NEUTRON UNIFORM 42GWD/2.55%/30.0Y NORMAL RUN 2010 c

c 10 20 30 40 50 60 70 c l l l l l l l c 3456789012345678901234567890123456789012345678901234567890123456789012345678 c

c ** This model changes the DBF per Prairie Island comments **

c ** Subcritical fission using MCNP **

c ** This model corrects composition errors in the homogenized Al/resin **

c ** This model credits boral modeled as aluminum at near full boral dens **

c ** This model corrects volume error in F4 tallies **

c ** This model corrects trunnion surface tallies F2/F12 **

c ** This model changes axial profile to assumed Prairie Island shape **

c c The following general design parameters are applied throughout this model:

c c 1. Uniform loading (modeled as two zones) c 2. Homogenized fuel; discrete steel box and basket aluminm c 3. Trunnions and impact limiters included c 4. Assumed burnup profile based on PI plant data c 5. Generic neutron source spectrum using builtin MCNP spectrum c 6. 10CFR71 acceptance criteria for dose locations c 7. Approximated periphery basket (including rails and outer SST inserts) c 8. Fuel enrichment 3.0 wt. % assumed c 9. Resin boxes modeled as homogenized Al box & resin mix c 10.Conservative Boral credit modeled as aluminum around fuel tubes c

c c ==============================================================================

c = =

c = CELL CARDS =

c = =

c ==============================================================================

c 1 0 -11:+12:+13 imp:p=0.0E+00 imp:n=0.0E+00 c

c fuel lattice c

c center fuel assys 2011 10 -2.45 (-275:+276:-277:+278) u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

boral gap 2021 0 (-295:+296:-297:+298) &

+211 -212 +213 -214 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

void around fuel 2031 6 -7.940 (-211:+212:-213:+214) &

+271 -272 +273 -274 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

steel liner 2041 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

aluminum sandwich 2061 0 +295 -296 +297 -298 -205 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

void gap 2051 4 -2.595 +295 -296 +297 -298 +205 -208 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

bottom end fitting 2071 1 -3.949 +295 -296 +297 -298 +208 -250 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

fuel 2081 2 -1.543 +295 -296 +297 -298 +250 -255 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

plenum 2091 3 -1.970 +295 -296 +297 -298 +255 -260 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

top end fitting Page 5-20

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2101 0 +295 -296 +297 -298 +260 u=11 imp:p=1.0E+00 imp:n=1.0E+00 $

void gap c

c periphery fuel assys 2012 10 -2.45 (-275:+276:-277:+278) u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

boral gap 2022 0 (-295:+296:-297:+298) &

+211 -212 +213 -214 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

void around fuel 2032 6 -7.940 (-211:+212:-213:+214) &

+271 -272 +273 -274 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

steel liner 2042 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

aluminum sandwich 2052 4 -2.595 +295 -296 +297 -298 -208 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

bottom end fitting 2062 0 +295 -296 +297 -298 -205 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

void gap 2072 1 -3.949 +295 -296 +297 -298 +208 -250 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

fuel 2082 2 -1.543 +295 -296 +297 -298 +250 -255 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

plenum 2092 3 -1.970 +295 -296 +297 -298 +255 -260 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

top end fitting 2102 0 +295 -296 +297 -298 +260 u=12 imp:p=1.0E+00 imp:n=1.0E+00 $

void gap c

c c empty compartment and periphery basket 2013 10 -2.702 (-275:+276:-277:+278) u=2 imp:p=1.0E+00 imp:n=1.0E+00 $

boral gap 2043 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=2 imp:p=1.0E+00 imp:n=1.0E+00 $

aluminum sandwich 2073 0 +271 -272 +273 -274 u=2 imp:p=1.0E+00 imp:n=1.0E+00 $

empty basket 241 10 -2.702 +272:-273 u=5 imp:p=1.0E+00 imp:n=1.0E+00 $

solid aluminum 245 0 -272 +273 u=5 imp:p=1.0E+00 imp:n=1.0E+00 $

empty assy 251 10 -2.702 -271:-273 u=6 imp:p=1.0E+00 imp:n=1.0E+00 $

basket 255 0 +271 +273 u=6 imp:p=1.0E+00 imp:n=1.0E+00 $

empty assy 261 10 -2.702 -271:+274 u=7 imp:p=1.0E+00 imp:n=1.0E+00 $

basket 265 0 +271 -274 u=7 imp:p=1.0E+00 imp:n=1.0E+00 $

empty assy 271 10 -2.702 +272:+274 u=8 imp:p=1.0E+00 imp:n=1.0E+00 $

basket 275 0 -272 -274 u=8 imp:p=1.0E+00 imp:n=1.0E+00 $

empty assy 299 0 -292 +291 -294 +293 u=9 imp:p=1.0E+00 imp:n=1.0E+00 &

lat=1 fill=-1:8 -5:5 0:0 c

c x coords c -1 0 1 2 3 4 5 6 7 8 c

9 9 9 9 9 9 9 9 9 9 $ -5 9 9 9 2 2 2 2 9 9 9 $ -4 9 7 8 12 12 12 12 7 8 9 $ -3 9 2 12 11 11 11 11 12 2 9 $ -2 Page 5-21

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 9 2 12 11 11 11 11 12 2 9 $ -1 9 12 11 11 11 11 11 11 12 9 $ 0 9 2 12 11 11 11 11 12 2 9 $ 1 9 2 12 11 11 11 11 12 2 9 $ 2 9 6 5 12 12 12 12 6 5 9 $ 3 9 9 9 2 2 2 2 9 9 9 $ 4 9 9 9 9 9 9 9 9 9 9 $ 5 c

c cavity cells 23 0 +201 -261 -2990 fill=9 imp:p=1.0E+00 imp:n=1.0E+00 24 6 -7.940 +201 -261 +2990 -299 imp:p=1.0E+00 imp:n=1.0E+00 $ outer SST inserts 2771 10 -2.702 +201 -261 +299 -282 -2991 imp:p=1.0E+00 imp:n=1.0E+00 $ these are the rails 2772 10 -2.702 +201 -261 +299 -282 +2991 -2992 imp:p=1.0E+00 imp:n=1.0E+00 2773 10 -2.702 +201 -261 +299 -282 +2992 -2993 imp:p=1.0E+00 imp:n=1.0E+00 2774 10 -2.702 +201 -261 +299 -282 +2993 -2994 imp:p=1.0E+00 imp:n=1.0E+00 2775 10 -2.702 +201 -261 +299 -282 +2994 -2995 imp:p=1.0E+00 imp:n=1.0E+00 2776 10 -2.702 +201 -261 +299 -282 +2995 -2996 imp:p=1.0E+00 imp:n=1.0E+00 2778 10 -2.702 +201 -261 +299 -282 +2996 imp:p=1.0E+00 imp:n=1.0E+00 c

c main shield shell cells 301 5 -7.702 (261 -511 -801):&

(802 801 282 -311):(419 -201 -802) imp:p=1.0E+00 imp:n=1.0E+00 302 5 -7.702 (511 -512 -801):&

(802 801 311 -312):(418 -419 -802) imp:p=1.0E+00 imp:n=1.0E+00 303 5 -7.702 (512 -513 -801):&

(802 801 312 -313):(417 -418 -802) imp:p=1.0E+00 imp:n=1.0E+00 304 5 -7.702 (513 -514 -801):&

(802 801 313 -314):(416 -417 -802) imp:p=1.0E+00 imp:n=1.0E+00 305 5 -7.702 (514 -515 -801):&

(802 801 314 -315):(415 -416 -802) imp:p=1.0E+00 imp:n=1.0E+00 306 5 -7.702 (515 -516 -801):&

(802 -516 315 -316):(414 -415 -802) imp:p=1.0E+00 imp:n=1.0E+00 307 5 -7.702 (802 -516 316 -317):(413 -414 -802) imp:p=1.0E+00 imp:n=1.0E+00 308 5 -7.702 (802 -516 317 -318):(412 -413 -802) imp:p=1.0E+00 imp:n=1.0E+00 309 5 -7.702 (802 -516 318 -319):(411 -412 -802) imp:p=1.0E+00 imp:n=1.0E+00 310 5 -7.702 (802 -516 319 -320):(410 -411 -802) imp:p=1.0E+00 imp:n=1.0E+00 c

c top of lid 501 5 -7.702 +516 -517 -521 imp:p=1.0E+00 imp:n=1.0E+00 502 5 -7.702 +517 -518 -521 imp:p=1.0E+00 imp:n=1.0E+00 503 5 -7.702 +518 -519 -521 imp:p=1.0E+00 imp:n=1.0E+00 504 5 -7.702 +519 -520 -521 imp:p=3.0E+00 imp:n=3.0E+00 505 10 -2.702 +520 -522 -320 imp:p=3.0E+00 imp:n=3.0E+00 $ Al impact limiter spacer on top c

c top impact limiters 601 0 (+622 -750 +320 -651) imp:p=1.0E+00 imp:n=1.0E+00 602 0 (+622 -750 +651 -672) imp:p=1.0E+00 imp:n=1.0E+00 603 8 -.125 (+750 -752 +320 -651) imp:p=1.0E+00 imp:n=1.0E+00 604 8 -.125 (+750 -752 +651 -672) imp:p=1.0E+00 imp:n=1.0E+00 605 10 -2.702 (+516 -518 +521 -318) imp:p=1.0E+00 imp:n=1.0E+00 606 10 -2.702 (+518 -520 +521 -318):&

(+516 -520 +318 -320) imp:p=3.0E+00 imp:n=3.0E+00 607 9 -.387 (+522 -721 -651):&

(+752 -522 +320 -651) imp:p=3.9E+00 imp:n=3.9E+00 608 9 -.387 (+752 -721 +651 -672):&

(+721 -722 -672) imp:p=4.2E+00 imp:n=4.2E+00 609 9 -.387 +722 -723 -672 imp:p=4.4E+00 imp:n=4.4E+00 610 9 -.387 +723 -724 -672 imp:p=4.1E+00 imp:n=4.1E+00 611 9 -.387 +724 -725 -672 imp:p=4.5E+00 imp:n=4.5E+00 612 9 -.387 +725 -726 -672 imp:p=5.1E+00 imp:n=5.1E+00 Page 5-22

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 613 9 -.387 +726 -727 -672 imp:p=6.5E+00 imp:n=6.5E+00 614 9 -.387 +727 -728 -672 imp:p=8.4E+00 imp:n=8.4E+00 615 9 -.387 +728 -729 -672 imp:p=1.0E+01 imp:n=1.0E+01 616 9 -.387 +729 -730 -672 imp:p=1.3E+01 imp:n=1.3E+01 617 9 -.387 +730 -731 -672 imp:p=1.5E+01 imp:n=1.5E+01 618 9 -.387 +731 -732 -672 imp:p=2.0E+01 imp:n=2.0E+01 619 9 -.387 +732 -733 -672 imp:p=2.8E+01 imp:n=2.8E+01 620 9 -.387 +733 -734 -672 imp:p=4.0E+01 imp:n=4.0E+01 621 9 -.387 +734 -735 -672 imp:p=6.5E+01 imp:n=6.5E+01 622 8 -.125 +735 -762 -672 imp:p=1.0E+02 imp:n=1.0E+02 623 6 -7.940 +762 -736 -672 imp:p=1.0E+02 imp:n=1.0E+02 c

632 0 +622 -750 +651 +672 -810 imp:p=1.1E+01 imp:n=1.1E+01 634 8 -.125 +750 -752 +651 +672 -810 imp:p=6.3E+00 imp:n=6.3E+00 638 9 -.387 +752 -723 +651 +672 -810 imp:p=5.5E+00 imp:n=5.5E+00 640 9 -.387 +723 -725 +672 -810 imp:p=6.2E+00 imp:n=6.2E+00 642 9 -.387 +725 -727 +672 -810 imp:p=9.4E+00 imp:n=9.4E+00 644 9 -.387 +727 -729 +672 -810 imp:p=1.4E+01 imp:n=1.4E+01 646 9 -.387 +729 -731 +672 -810 imp:p=2.2E+01 imp:n=2.2E+01 648 9 -.387 +731 -733 +672 -810 imp:p=4.0E+01 imp:n=4.0E+01 650 9 -.387 +733 -735 +672 -810 imp:p=9.1E+01 imp:n=9.1E+01 652 8 -.125 +735 -762 +672 -810 imp:p=2.4E+02 imp:n=2.4E+02 653 6 -7.940 +762 -736 +672 -810 imp:p=2.4E+02 imp:n=2.4E+02 c

662 0 +622 -750 +651 +810 -811 imp:p=2.4E+01 imp:n=2.4E+01 664 8 -.125 +750 -752 +651 +810 -811 imp:p=1.5E+01 imp:n=1.5E+01 668 9 -.387 +752 -723 +651 +810 -811 imp:p=1.2E+01 imp:n=1.2E+01 672 9 -.387 +723 -727 +810 -811 imp:p=1.7E+01 imp:n=1.7E+01 676 9 -.387 +727 -731 +810 -811 imp:p=3.2E+01 imp:n=3.2E+01 680 9 -.387 +731 -735 +810 -811 imp:p=1.2E+02 imp:n=1.2E+02 682 8 -.125 +735 -762 +810 -811 imp:p=6.2E+02 imp:n=6.2E+02 683 6 -7.940 +762 -736 +810 -811 imp:p=6.2E+02 imp:n=6.2E+02 c

c side neutron shield 851 7 -1.642 +905 -904 +650 -651 +861 imp:p=1.0E+00 imp:n=1.0E+00 852 7 -1.642 +905 -904 +651 -670 +861 imp:p=1.0E+00 imp:n=1.0E+00 853 5 -7.702 +621 -622 +320 -671 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 854 5 -7.702 +601 -602 +320 -671 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 855 5 -7.702 +905 -904 +671 -672 +861 imp:p=1.0E+00 imp:n=1.0E+00 856 7 -1.642 +905 -904 +320 -650 +861 imp:p=1.0E+00 imp:n=1.0E+00 857 7 -1.642 +905 -904 +670 -671 +861 imp:p=1.0E+00 imp:n=1.0E+00 c Neutron shielding - Flat - Top 858 7 -1.642 +900 -621 +650 -651 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 859 7 -1.642 +900 -621 +651 -670 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 860 7 -1.642 +900 -621 +320 -650 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 861 7 -1.642 +900 -621 +670 -671 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 862 7 -1.642 +904 -900 +650 -651 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 863 7 -1.642 +904 -900 +651 -670 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 864 7 -1.642 +904 -900 +320 -650 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 865 7 -1.642 +904 -900 +670 -671 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 c Neutron shielding - Flat - Bottom 866 7 -1.642 +602 -901 +650 -651 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 867 7 -1.642 +602 -901 +651 -670 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 868 7 -1.642 +602 -901 +320 -650 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 869 7 -1.642 +602 -901 +670 -671 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 870 7 -1.642 +901 -905 +650 -651 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 871 7 -1.642 +901 -905 +651 -670 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 872 7 -1.642 +901 -905 +320 -650 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 873 7 -1.642 +901 -905 +670 -671 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 c

874 5 -7.702 +900 -622 +671 -672 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 875 5 -7.702 +904 -900 -907 -672 +861 imp:p=1.0E+00 imp:n=1.0E+00 876 5 -7.702 +900 +861 +903 -907 -622 -672 imp:p=1.0E+00 imp:n=1.0E+00 Page 5-23

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 877 5 -7.702 +900 +861 +906 -902 -622 -672 imp:p=1.0E+00 imp:n=1.0E+00 878 5 -7.702 +904 -900 +906 -672 +861 imp:p=1.0E+00 imp:n=1.0E+00 879 5 -7.702 +904 -900 +907 -906 -672 +671 imp:p=1.0E+00 imp:n=1.0E+00 880 5 -7.702 +601 -901 +671 -672 +861 -906 907 imp:p=1.0E+00 imp:n=1.0E+00 881 5 -7.702 +901 -905 -907 -672 +861 imp:p=1.0E+00 imp:n=1.0E+00 882 5 -7.702 +901 -905 +906 -672 +861 imp:p=1.0E+00 imp:n=1.0E+00 883 5 -7.702 +901 -905 +907 -906 -672 +671 imp:p=1.0E+00 imp:n=1.0E+00 884 5 -7.702 +601 +864 +903 -907 -901 -672 imp:p=1.0E+00 imp:n=1.0E+00 885 5 -7.702 +601 +864 +906 -902 -901 -672 imp:p=1.0E+00 imp:n=1.0E+00 c Air 886 0 +900 -622 +861 -903 -672 imp:p=1.0E+00 imp:n=1.0E+00 887 0 +900 -622 +861 +902 -672 -908 imp:p=1.0E+00 imp:n=1.0E+00 888 0 +900 -622 +861 +902 -672 +908 imp:p=1.0E+00 imp:n=1.0E+00 889 0 +601 -901 -903 -672 imp:p=1.0E+00 imp:n=1.0E+00 890 0 +601 -901 +902 -672 imp:p=1.0E+00 imp:n=1.0E+00 c bottom impact limiters c

701 8 -.125 (+751 -601 +320 -651 -875):&

(+751 -601 +320 -651 876) imp:p=1.0E+00 imp:n=1.0E+00 702 8 -.125 (+751 -601 +651 -672 -875):&

(+751 -601 +651 -672 876) imp:p=1.0E+00 imp:n=1.0E+00 7031 9 -.387 (+410 -751 +320 -651 ) &

(-875:876:-879) imp:p=1.0E+00 imp:n=1.0E+00 7032 9 -.387 (+716 -410 -651) imp:p=2.3E+00 imp:n=2.3E+00 7041 9 -.387 (+716 -751 +651 -672) &

(-875:876:-879) imp:p=1.0E+00 imp:n=1.0E+00 7042 9 -.387 (+715 -716 -672) imp:p=2.5E+00 imp:n=2.5E+00 705 9 -.387 +714 -715 -672 imp:p=2.5E+00 imp:n=2.5E+00 706 9 -.387 +713 -714 -672 imp:p=2.8E+00 imp:n=2.8E+00 707 9 -.387 +712 -713 -672 imp:p=3.2E+00 imp:n=3.2E+00 708 9 -.387 +711 -712 -672 imp:p=3.4E+00 imp:n=3.4E+00 709 9 -.387 +710 -711 -672 imp:p=4.0E+00 imp:n=4.0E+00 710 9 -.387 +709 -710 -672 imp:p=5.0E+00 imp:n=5.0E+00 711 9 -.387 +708 -709 -672 imp:p=6.5E+00 imp:n=6.5E+00 712 9 -.387 +707 -708 -672 imp:p=8.0E+00 imp:n=8.0E+00 713 9 -.387 +706 -707 -672 imp:p=9.8E+00 imp:n=9.8E+00 714 9 -.387 +705 -706 -672 imp:p=1.4E+01 imp:n=1.4E+01 715 9 -.387 +704 -705 -672 imp:p=1.9E+01 imp:n=1.9E+01 716 9 -.387 +703 -704 -672 imp:p=2.9E+01 imp:n=2.9E+01 717 9 -.387 +761 -703 -672 imp:p=2.9E+01 imp:n=2.9E+01 718 8 -.125 +702 -761 -672 imp:p=6.7E+01 imp:n=6.7E+01 719 6 -7.940 +701 -702 -672 imp:p=6.7E+01 imp:n=6.7E+01 c

721 0 +753 -601 +672 -810 &

(-875:876:-877:878) imp:p=5.4E+00 imp:n=5.4E+00 722 8 -.125 +751 -753 +672 -810 &

(-875:876:-877:878) imp:p=5.4E+00 imp:n=5.4E+00 724 9 -.387 +715 -751 +672 -810 &

(-875:876:-877:878:-879) imp:p=4.3E+00 imp:n=4.3E+00 726 9 -.387 +713 -715 +672 -810 imp:p=4.5E+00 imp:n=4.5E+00 728 9 -.387 +711 -713 +672 -810 imp:p=6.1E+00 imp:n=6.1E+00 730 9 -.387 +709 -711 +672 -810 imp:p=9.5E+00 imp:n=9.5E+00 732 9 -.387 +707 -709 +672 -810 imp:p=1.6E+01 imp:n=1.6E+01 734 9 -.387 +705 -707 +672 -810 imp:p=2.3E+01 imp:n=2.3E+01 736 9 -.387 +761 -705 +672 -810 imp:p=5.2E+01 imp:n=5.2E+01 738 8 -.125 +702 -761 +672 -810 imp:p=1.9E+02 imp:n=1.9E+02 739 6 -7.940 +701 -702 +672 -810 imp:p=1.9E+02 imp:n=1.9E+02 c

741 0 +753 -601 +810 -811 imp:p=1.4E+01 imp:n=1.4E+01 742 8 -.125 +751 -753 +810 -811 imp:p=1.4E+01 imp:n=1.4E+01 744 9 -.387 +715 -751 +810 -811 imp:p=1.1E+01 imp:n=1.1E+01 748 9 -.387 +711 -715 +810 -811 imp:p=1.3E+01 imp:n=1.3E+01 752 9 -.387 +707 -711 +810 -811 imp:p=2.4E+01 imp:n=2.4E+01 Page 5-24

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 756 9 -.387 +761 -707 +810 -811 imp:p=8.0E+01 imp:n=8.0E+01 758 8 -.125 +702 -761 +810 -811 imp:p=5.4E+02 imp:n=5.4E+02 759 6 -7.940 +701 -702 +810 -811 imp:p=5.4E+02 imp:n=5.4E+02 c

760 0 (864 867 875 -876 320 -878 879 -601):&

(864 -867 875 -876 320 +877 879 -601) imp:p=1.0E+00 imp:n=1.0E+00 c

c tally cells 900 0 +601 -622 +672 -899 +861 imp:p=4.0E+00 imp:n=4.0E+00 901 0 +601 -622 +899 -810 +861 imp:p=4.0E+00 imp:n=4.0E+00 902 0 +601 -622 +810 -811 imp:p=1.2E+01 imp:n=1.2E+01 903 11 -1.225E-3 +701 -736 +811 -812 imp:p=1.2E+02 imp:n=1.2E+02 $

comment these out for importance optimization 904 11 -1.225E-3 (+701 -736 +812 -814):&

(+736 -815 -814):(+816 -701 -814) imp:p=6.2E+02 imp:n=6.2E+02 905 11 -1.225E-3 (+816 -815 +814 -813):&

(+815 -817 -813):(+818 -816 -813) imp:p=6.2E+02 imp:n=6.2E+02 906 11 -1.225E-3 (+11 13) (860:+817:-818) &

(+818:-11:810) (-817:12:810) imp:p=6.2E+02 imp:n=6.2E+02 907 11 -1.225E-3 +851 -818 -810 imp:p=6.2E+02 imp:n=6.2E+02 $ far cells at end of rail car 908 11 -1.225E-3 +850 -851 -810 imp:p=6.2E+02 imp:n=6.2E+02 909 11 -1.225E-3 +11 -850 -810 imp:p=6.2E+02 imp:n=6.2E+02 910 11 -1.225E-3 +817 -852 -810 imp:p=6.2E+02 imp:n=6.2E+02 911 11 -1.225E-3 +852 -853 -810 imp:p=6.2E+02 imp:n=6.2E+02 912 11 -1.225E-3 +853 810 imp:p=6.2E+02 imp:n=6.2E+02 913 11 -1.225E-3 (+11 860) (813:+817:-818) &

(+818:-11:810) (-817:12:810) imp:p=6.2E+02 imp:n=6.2E+02 c

c upper trunnions 801 5 -7.702 (-867 +320 +870 -861):&

(+867 +320 -871 -861) imp:p=1.0E+00 imp:n=1.0E+00 802 5 -7.702 (+869 -870 +863 -861):&

(+871 -872 +863 -861) imp:p=1.0E+00 imp:n=1.0E+00 803 5 -7.702 (+868 -869 +863 -862):&

(+872 -873 +863 -862) imp:p=1.0E+00 imp:n=1.0E+00 804 7 -1.642 (+868 -870 -863):&

(+871 -873 -863) imp:p=1.0E+00 imp:n=1.0E+00 805 0 (-869 -810 -861 (+862:-868)):&

(+872 -810 -861 (+862:+873)) imp:p=1.0E+00 imp:n=1.0E+00 c

c lower trunnions 811 5 -7.702 (-867 +320 +911 -864):&

(+867 +320 -912 -864) imp:p=4.0E+00 imp:n=4.0E+00 812 5 -7.702 (+869 -911 +866 -864):&

(+912 -872 +866 -864) imp:p=2.5E+00 imp:n=2.5E+00 813 5 -7.702 (+868 -869 +866 -865):&

(+872 -873 +866 -865) imp:p=5.0E+00 imp:n=5.0E+00 814 0 (+868 -911 -866):&

(+912 -873 -866) imp:p=5.4E+00 imp:n=5.4E+00 815 0 (+877 -869 +865 -864):&

(+872 -878 +865 -864) imp:p=5.0E+00 imp:n=5.0E+00 816 0 ( +877 -868 -865):&

(+873 -878 -865) imp:p=5.4E+00 imp:n=5.4E+00 c ==============================================================================

c = =

c = SURFACE CARDS =

c = =

c ==============================================================================

c c model boundaries c initial model had rail car at 41' 8" long Page 5-25

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c rail cars typically are 40 to 60 feet c PacTec typical transport car (for OCRWM) 49' 6" c set outer bounds at 60 feet c assume cask CG is middle of active fuel (within 6 inches of actual) 11 pz -914.4 $ end of rail car 12 pz 914.4 $ end of rail car 13 cz 652.4 $ 5 meter from (vertical planes) side of rail car c

c fuel/basket region 201 pz -190.70 $ bottom of lower end fitting (bottom cavity) 205 pz -190.7032 $ top of lower end fitting (3.08")

208 pz -182.88 $ bottom of active fuel (144")

211 px -88.7034 $ steel liner inner (8.05 inch typ. square) 212 px -68.20565 $ steel liner modeled as 13 ga. 0.09" thick not 0.10" 213 py -10.2494 $

214 py 10.2494 $

250 pz 182.88 $ top of active fuel 255 pz 201.02068 $ top of plenum (7.142")

260 pz 209.91 $ top of upper end fitting (3.5")

261 pz 223.32 $ top of cavity (void) 271 px -88.93175 $ steel liner outer (8.05 + .2 inch square) 272 px -67.97675 $

273 py -10.4775 $

274 py 10.4775 $

275 px -89.56675 $ boral gap (8.05 + .7 inch square) 276 px -67.34175 $

277 py -11.1125 $

278 py 11.1125 $

282 cz 91.44 $ inner radius of cavity (72 inch diameter) 291 px -89.662 $ basket lattice unit cell (8.05+.775 inch square) 292 px -67.2465 $ origin (center of assy) @ x=-78.45425 y=0 293 py -11.20775 $

294 py 11.20775 $

295 px -88.31326 $ fuel envelope 7.763 x 7.763 inch 296 px -68.59524 $

297 py -9.85901 $

298 py 9.85901 $

299 cz 90.932 $ assumes .2" instead of 3/8" outer basket ("rails") to conserve mass 2990 cz 90.867 $ assumes 0.026" for steel inserts on periphery 2991 py -79.4385 $ rail cutting surfaces 2992 py -55.0545 $

2993 py -12.192 $

2994 py 12.192 $

2995 py 55.0545 $

2996 py 79.4385 $

c c wall gamma shield 311 cz 93.853 $ split primary gamma shield into 10 layers 312 cz 96.266 $ layer thickness = 2.413 cm or 0.95 inch 313 cz 98.679 $ also applies to bottom and top (diff. thickness) 314 cz 101.092 $

315 cz 103.505 $

316 cz 105.918 $

317 cz 108.331 $

318 cz 110.744 $

319 cz 113.157 $

320 cz 115.57 $ outside cask body c

c bottom gamma shield 410 pz -216.73 $ cask bottom 411 pz -214.127 $ layer thickness = 2.603 cm or 1.025 inch 412 pz -211.524 $

Page 5-26

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 413 pz -208.921 $

414 pz -206.318 $

415 pz -203.715 $

416 pz -201.112 $

417 pz -198.509 $

418 pz -195.906 $

419 pz -193.303 $

c c top gamma shield 511 pz 225.987 $ layer thickness = 2.667 cm or 1.05 inch 512 pz 228.654 $

513 pz 231.321 $

514 pz 233.988 $

515 pz 236.655 $

516 pz 238.56 $ top of cask walls 517 pz 241.989 $

518 pz 244.656 $

519 pz 247.323 $

520 pz 249.99 $

521 cz 105.09 $ outer radius of top part of lid 522 pz 252.022 $ assume 0.8" thick to account for tolerences and bolt holes c

c neutron shields c 601 pz -186.25 $ bottom edge of side resin lower housing 601 pz -185.615 $ bottom edge of side resin lower housing c 602 pz -184.34 $ top edge of side resin lower housing 602 pz -183.705 $ top edge of side resin lower housing 621 pz 203.00 $ bottom edge of side resin upper housing 622 pz 204.91 $ top edge of side resin upper housing 650 cz 116.1542 $ homogenized inner aluminum box 651 cz 121.285 $ splitting shell for neutron shielding 670 cz 126.4158 $ homogenized outer aluminum box 671 cz 127.0 $ outer radius of resin shield on side 672 cz 128.27 $ outer radius of resin housing on side c

c impact limiters 701 pz -311.98 702 pz -311.345 703 pz -299.05 704 pz -293.17 705 pz -287.29 706 pz -281.41 707 pz -275.53 708 pz -269.65 709 pz -263.77 710 pz -257.89 711 pz -252.01 712 pz -246.13 713 pz -240.25 714 pz -234.37 715 pz -228.49 716 pz -222.61 c

721 pz 255.87 722 pz 261.75 723 pz 267.63 724 pz 273.51 725 pz 279.39 726 pz 285.27 727 pz 291.15 728 pz 297.03 729 pz 302.91 Page 5-27

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 730 pz 308.79 731 pz 314.67 732 pz 320.55 733 pz 326.43 734 pz 332.31 735 pz 338.255 736 pz 345.24 c

750 pz 220.78 $ bottom of top impact limiter 751 pz -195.14 $ balsa/redwood boundary 752 pz 228.4 $ balsa/redwood boundary 753 pz -187.52 $ balsa/redwood boundary 761 pz -304.995 $ balsa/redwood boundary 762 pz 344.605 $ balsa/redwood boundary c

c miscellaneous surfaces 801 kz 122.2547369 0.8185941 +1 $ upper splitting mating plane 802 kz -92.06 0.8593425329 -1 $ lower splitting mating plane 899 cz 128.271 $ void cell outside cask 810 cz 152.4 $ vertical planes on side (10' wide rail car) 811 cz 182.88 $ top/bottom @ 12' diameter impact limiters 812 cz 198.12 $ top/bottom @ 13' diameter impact limiters (not used) 813 cz 352.4 $ 2 meter from vertical planes on side 860 cz 383.0 $ 2 meter from vertical impact limiter 814 cz 215.57 $ 1 meter from cask body (HAC) 815 pz 349.99 $ 1 meter from cask top lid (HAC) 816 pz -316.73 $ 1 meter from cask bottom (HAC) 817 pz 545.24 $ 2 meter from top impact limiter 818 pz -511.98 $ 2 meter from bottom impact limiter 824 pz -190 $ axial tally segments 825 pz -170 $ 20 cm wide 826 pz -150 $ centered on fuel midplane 827 pz -130 828 pz -110 829 pz -90 830 pz -70 831 pz -50 832 pz -30 833 pz -10 834 pz 10 835 pz 30 836 pz 50 837 pz 70 838 pz 90 839 pz 110 840 pz 130 841 pz 150 842 pz 170 843 pz 190 844 pz 210 845 pz 230 846 pz 250 c occupied position surfaces 850 pz -762.0 $ 40 foot rail car 851 pz -609.6 $ 50 foot rail car 852 pz 609.6 853 pz 762.0 882 cz 30 $ radial segmentation 883 cz 55 885 cz 140.57 886 cz 165.57 887 cz 232.88 888 cz 282.88 Page 5-28

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 861 c/y 0 177.6 15.254 $ top trunnion base radius 862 c/y 0 177.6 14.2875 $ top trunnion 863 c/y 0 177.6 6.35 $ top trunnion resin radius 864 c/y 0 -197.685 12.0624 $ bottom trunnion base (9.498" dia.)

865 c/y 0 -197.685 11.2776 $ bottom trunnion 866 c/y 0 -197.685 5.715 $ bottom trunnion hole radius 867 py 0 $ trunnion ambiguious surface 868 py -132.715 $ NEW distance between trunnion head 104.5" 869 py -128.27 870 py -123.825 871 py 123.825 872 py 128.27 873 py 132.715 $ NEW distance between trunnion head 104.5" c

c lower trunnion "channel" in impact limiter 875 px -12.6974 $ these surfs form a box around lower trunnions 876 px 12.6974 $ 1/4" gap is assumed 877 py -136.8552 $

878 py 136.8552 $

879 pz -210.3824 $

c 900 pz 133.785 $ 17.25" from surf. center of top trunnion 901 pz -153.865 $ estimate 17.25" from surf. center of trunnion 902 py 124.7648 $ distance between surf. 902-903 flat -> 98.24" 903 py -124.7648 $

904 pz 132.515 905 pz -152.595 906 py 123.4948 $ 0.5" from surf. 902 907 py -123.4948 $ 0.5" from surf. 903 c

908 pz 177.6 $

c 911 py -122.047 912 py 122.047 c ==============================================================================

c = =

c = DATA CARDS =

c = =

c ==============================================================================

c c materials c

c **************************************************************

c In-Core Region c Density = 3.949 g/cm^3; Composition by weight fraction c 3.0 w/o enrichment c 410 kg U c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m1 92234 -0.000195 $ U-234 92235 -0.021902 $ U-235 92236 -0.000101 $ U-236 92238 -0.707883 $ U-238 8016 -0.098148 $ O 6012 -0.000011 $ C 14000 -0.000377 $ Si 15031 -0.000006 $ P 24000 -0.004200 $ Cr 22000 -0.000239 $ Ti 25055 -0.000276 $ Mn 26000 -0.010400 $ Fe Page 5-29

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 28000 -0.008289 $ Ni 40000 -0.145805 $ Zr 50000 -0.002152 $ Sn 72000 -0.000015 $ Hf c 1001 0.000000 $ H c

c **************************************************************

c Plenum c Density = 1.543 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m2 6012 -0.00045 $ C 14000 -0.00714 $ Si 15031 -0.00025 $ P 24000 -0.11569 $ Cr 22000 -0.00156 $ Ti 25055 -0.01115 $ Mn 26000 -0.38641 $ Fe 28000 -0.09859 $ Ni 40000 -0.37321 $ Zr 50000 -0.00551 $ Sn 72000 -0.00004 $ Hf c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Top Region c Density = 1.970 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m3 6012 -0.00074 $ C 14000 -0.01112 $ Si 15031 -0.00042 $ P 24000 -0.18702 $ Cr 22000 -0.00187 $ Ti 25055 -0.01851 $ Mn 26000 -0.63795 $ Fe 28000 -0.14238 $ Ni c 40000 0.00000 $ Zr c 50000 0.00000 $ Sn c 72000 0.00000 $ Hf c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Bottom Region c Density = 2.595 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m4 6012 -0.00080 $ C 14000 -0.01000 $ Si 15031 -0.00045 $ P 24000 -0.19000 $ Cr c 22000 0.00000 $ Ti 25055 -0.02000 $ Mn 26000 -0.68375 $ Fe 28000 -0.09500 $ Ni c 40000 0.00000 $ Zr c 50000 0.00000 $ Sn c 72000 0.00000 $ Hf Page 5-30

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Carbon Steel c Density = 7.8212 g/cm^3 SCALE Standard Comp. Library --> 98.47%

c **************************************************************

c m5 6012 -0.0100 $ C 26000 -0.9900 $ Fe c

c **************************************************************

c Stainless Steel 304 c Density = 7.94 g/cm^3 SCALE Standard Comp. Library c **************************************************************

c m6 26000 -0.68375 $ Fe 24000 -0.19000 $ Cr 28000 -0.09500 $ Ni 25055 -0.02000 $ Mn 14000 -0.01000 $ Si 6012 -0.00080 $ C 15031 -0.00045 $ P c

c **************************************************************

c Homogenized Neutron Resin/Aluminum Shield c

Reference:

Page 6, TN Calc. 1042-08, Rev. 0 c Based on TN-24 Resin c Density = 1.69 g/cm^3 --> 97.2%

c B-10 and B-11 based on natural abundance c Note Al is in resin and the aluminum boxes c **************************************************************

c m7 1001 -0.0418 $ H-1 5010 -0.0016 $ B-10 5011 -0.0071 $ B-11 6000 -0.2908 $ C 8016 -0.3455 $ O 13027 -0.2958 $ Al 30000 -0.0175 $ Zn c

c **************************************************************

c Balsa Wood c TN-68 SAR c Density = 0.125 g/cm^3 c **************************************************************

c m8 6012 0.2857 $ C 8016 0.2381 $ O 1001 0.4762 $ H c

c **************************************************************

c Redwood c Assume same compositin as Balsa c TN-68 SAR c Density = 0.387 g/cm^3 c **************************************************************

c m9 6012 0.2857 $ C 8016 0.2381 $ O 1001 0.4762 $ H c

c **************************************************************

Page 5-31

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c Pure Aluminum c Density = 2.702 g/cm^3 SCALE Standard Comp. Library c **************************************************************

c m10 13027 -1.0 $ Al c

c **************************************************************

c AIR: ANSI/ANS-6.6.1 Dry air c Density = 0.001225 g/cm^3 c Composition by weight fraction c **************************************************************

c m11 6012 -0.00014 $ C 7014 -0.75519 $ N 8016 -0.23179 $ O 18000 -0.01288 $ Ar c

c **************************************************************

c BORAL: TN40 STORAGE CRITICALITY SPECIFICATION c Density = 2.45 g/cm^3 (770 lbs of boral, about 95.8% theo.)

c Composition by weight fraction c Based on Calculation 1042-6 (TN40 Storage Crit) c 0.01 gm/cm^2 B-10 areal density c core thickness of 0.025" c panel thickness of 0.075" c redistribution of wt. frac. to simulate 75% B4C c **************************************************************

c actual wt. frac m12 6012 -0.066 $ 0.088 13027 -0.695 $ 0.594 5010 -0.044 $ 0.058 5011 -0.195 $ 0.060 c source c

sdef cel=d1 x=d2 y=d3 z=fcel=d4 erg=d5 par=1 c

c 22 inner assemblies (0.55) c 18 outer assemblies (0.45) c c inner outer si1 s 11 12 $ use distribution numbers to separate zones sp1 0.550 0.450 $ sample based on source strength or num assy if uniform c

c INNER ASSYS OUTER ASSYS

si11 sp11 si12 sp12 l d l d $

23:299:2071 1 23:299:2072 1 $ fuel zone c

c sample source/fuel cell c sample volume uniformly c use lattice element (0,0) c c X DIMENSIONS Y DIMENSIONS

si2 sp2 si3 sp3

-88.31326 0 -9.85901 0

-68.59524 1 9.85901 1 c

c zone dependent axial distributions c burnup is taken from Prairie site specific calculation c cask center c

ds4 s 41 42 $ distribution numbers c

Page 5-32

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c INNER ASSYS OUTER ASSYS

si41 sp41 si42 sp42

-182.88 0 -182.88 0

-152.4 0.0146 -152.4 0.0146

-121.92 0.0686 -121.92 0.0686

-91.44 0.0959 -91.44 0.0959

-60.96 0.1104 -60.96 0.1104

-30.48 0.1104 -30.48 0.1104 0 0.1104 0 0.1104 30.48 0.1096 30.48 0.1096 60.96 0.1066 60.96 0.1066 91.44 0.1044 91.44 0.1044 121.92 0.0959 121.92 0.0959 152.4 0.0586 152.4 0.0586 182.88 0.0146 182.88 0.0146 c

c generic Cm-244 fission spectrum c

sp5 -3 0.906 3.848 c

c following is old storage spectrum c ds5 s 51 52 $ distribution numbers c

c INNER ASSYS OUTER ASSYS c # si51 sp51 si52 sp52 c .100 0.00000 .100 0.00000 c .400 0.03627 .400 0.03627 c .900 0.18538 .900 0.18538 c 1.40 0.17223 1.40 0.17223 c 1.85 0.13081 1.85 0.13081 c 3.00 0.24536 3.00 0.24536 c 6.434 0.21231 6.434 0.21231 c 20 0.01764 20 0.0176 c

c surface tallies c

c fc2 NEUTRON TALLY ON TRUNNION SURFACE f2:n 868 873 fs2 -862 NT sd2 641 1E-10 641 1E-10 de2 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df2 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc4 NEUTRON VOLUME TALLIES AROUND THE TRUNNION f4:n 601 602 632 662 805 sd4 4j 2.4E4 de4 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df4 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc12 GAMMA TALLY ON TRUNNION SURFACE f12:p 868 873 fs12 -862 NT sd12 641 1E-10 641 1E-10 de12 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 Page 5-33

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df12 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc14 GAMMA VOLUME TALLIES AROUND THE TRUNNION f14:p 601 602 632 662 805 sd14 4j 2.4E4 de14 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df14 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc102 NEUTRON TALLY: SIDE, TOP/BOTTOM D=12', TOP/BOTTOM D=13.5', 5M f102:n 672 810 811 812 813 860 13 NT fs102 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T sd102 94215 16119 17R 15883 14983 16123 16199 72664 520209 150J de102 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df102 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc112 NEUTRON TALLY: CONTACT AND 2 METERS FROM IMPACT LIMITERS f112:n 818 701 736 817 NT fs112 -882 -883 -282 -320 -885 -886 -811 T de112 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df112 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc122 NEUTRON TALLY: 1 METER SIDE ACCIDENT f122:n 814 fs122 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T de122 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df122 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc132 NEUTRON TALLY: 1 METER END ACCIDENT, BOTTOM, TOP f132:n 816 815 NT fs132 -882 -883 -282 -320 -885 -886 -811 T de132 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df132 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 Page 5-34

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c

c fc142 NEUTRON TALLY: TOP RAIL CAR EDGE AS A FUNCTION OF DISTANCE f142:n 852 853 12 $ 40', 50', 60' rail car de142 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df142 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc152 NEUTRON TALLY: BOTTOM RAIL CAR EDGE AS A FUNCTION OF DISTANCE f152:n 851 850 11 $ 40', 50', 60' rail car de152 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df152 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc202 GAMMA TALLY: SIDE, TOP/BOTTOM D=12', TOP/BOTTOM D=13.5', 5M f202:p 672 810 811 812 813 860 13 NT fs202 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T sd202 94215 16119 17R 15883 14983 16123 16199 72664 520209 150J de202 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df202 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc212 GAMMA TALLY: CONTACT AND 2 METERS FROM IMPACT LIMITERS f212:p 818 701 736 817 NT fs212 -882 -883 -282 -320 -885 -886 -811 T de212 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df212 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc222 GAMMA TALLY: 1 METER SIDE ACCIDENT f222:p 814 fs222 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T de222 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df222 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 Page 5-35

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc232 GAMMA TALLY: 1 METER END ACCIDENT, BOTTOM, TOP f232:p 816 815 NT fs232 -882 -883 -282 -320 -885 -886 -811 T de232 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df232 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc242 GAMMA TALLY: TOP RAIL CAR EDGE AS A FUNCTION OF DISTANCE f242:p 852 853 12 $ 40', 50', 60' rail car de242 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df242 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc252 GAMMA TALLY: BOTTOM RAIL CAR EDGE AS A FUNCTION OF DISTANCE f252:p 851 850 11 $ 40', 50', 60' rail car de252 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df252 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c f24:p 906 fc24 FARTHEST TALLY FOR WWG c

fc5 NEUTRON TALLY ON FLAT SURFACE f5:n 0. 183. 145. 2.

de5 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df5 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

fc505 GAMMA TALLY ON TRUNNION SURFACE f505:p 0. 183. 145. 2.

de505 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 Page 5-36

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df505 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

fc55 NEUTRON TALLY ON FLAT SURFACE 2m f55:n 0. 383. 145. 0.25 de55 LOG 2.5-8 1.0-7 1.0-6 1.0-5 1.0-4 1.0-3 1.0-2 1.0-1 5.0-1 1.0 2.5 5.0 7.0 10.0 14.0 20.0 df55 LOG 3.67-6 3.67-6 4.46-6 4.54-6 4.18-6 3.76-6 3.56-6 2.17-5 9.26-5 1.32-4 1.25-4 1.56-4 1.47-4 1.47-4 2.08-4 2.27-4 c

c fc515 GAMMA TALLY ON TRUNNION SURFACE 2m f515:p 0. 383. 145. 0.25 de515 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df515 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c control cards c

c nonu $ turn off subcritical fission mode p n $ coupled neutron, gamma mode c nps 2.0E7 $ cut nps at ~ ? hours ctme 7200 prdmp j j 1 2 j $ print once, dump every 15 min, MCTAL, keep last 2 dumps cut:p j 0.01 3j $ cut photons < 0.01 MeV bottom ANSI/ANS-6.1.1-1977 cut:n j 2.5E-8 3j $ cut neutron < 2.5E-8 MeV bottom ANSI/ANS-6.1.1-1977 phys:p 10 1 1 $ no bremsstrahlung, no coherent scattering for n, gammas print 10 $ Source coefficients and distribution

-20 $ Weight window information 30 $ Tally description

-35 $ Coincident detectors 40 $ Material composition 50 $ Cell volumes and masses, surface areas 60 $ basic Cell importances 62 $ basic Forced collision and exponential transform

-70 $ Surface coefficients

-72 $ basic Cell temperatures

-85 $ Electron range and straggling tables multigroup: flux values for biasing adjoint calcs

-86 $ Electron bremsstrahlung and secondary production

-90 $ KCODE source data

-98 $ Physical constants and compile options 100 $ basic Cross section tables

-102 $ Assignment of S(a,b) data to nuclides

-110 $ First 50 starting histories 120 $ Analysis of the quality of your importance function 126 $ basic Particle activity in each cell

-128 $ Universe map

-130 $ Neutron/photon/electron weight balance

-140 $ Neutron/photon nuclide activity

-150 $ DXTRAN diagnostics 160 $ default TFC bin tally analysis Page 5-37

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 161 $ default f(x) tally density plot 162 $ default Cumulative f(x) and tally density plot

-170 $ Source distribution frequency tables, surface source

-175 $ shorten Estimated keff results by cycle

-178 $ Estimated keff results by batch size

-180 $ Weight window generator bookkeeping summary controlled by WWG(7), not print card

-190 $ basic Weight window generator summary

-198 $ Weight windows from multigroup fluxes

-200 $ basic Weight window generated windows c

c below is the weight window c uncomment the following to generate the weight windows c wwg 24 2072 0 4J 0 MCNP Primary Gamma Input File TN-40 TRAN GAMMA UNIFORM 42GWD/2.55%/30.0Y NORMAL RUN 2010 c

c 10 20 30 40 50 60 70 c l l l l l l l c 3456789012345678901234567890123456789012345678901234567890123456789012345678 c

c ** This model changes the DBF per Prairie Island comments **

c ** This model corrects an error in the energy distribution of the TPAs **

c ** This model corrects composition errors in the homogenized Al/resin **

c ** This model extends TPA decay from 9 years (DB) to 13 years **

c ** This model credits boral modeled as aluminum at near full boral dens **

c ** This model corrects volume error in F4 tallies **

c ** This model corrects trunnion surface tallies F2/F12 **

c ** This model changes axial profile to assumed Prairie Island shape **

c c The following general design parameters are applied throughout this model:

c c 1. Uniform loading (modeled as two zones) c 2. Homogenized fuel; discrete steel box and basket aluminm c 3. Trunnions and impact limiters included c 4. Assumed burnup profile based on PI plant data c 5. Gamma source from DB Fuel (includes TPAs distributed in 2 groups) c 6. 10CFR71 acceptance criteria for dose locations c 7. Approximated periphery basket (including rails and outer SST inserts) c 8. Fuel enrichment 3.0 wt. % assumed c 9. Resin boxes modeled as homogenized Al box & resin mix c 10.Conservative Boral credit modeled as aluminum around fuel tubes c

c c ==============================================================================

c = =

c = CELL CARDS =

c = =

c ==============================================================================

c 1 0 -11:+12:+13 imp:p=0.0E+00 c

c fuel lattice c

c center fuel assys 2011 10 -2.45 (-275:+276:-277:+278) u=11 imp:p=1.0E+00 $ boral gap 2021 0 (-295:+296:-297:+298) &

+211 -212 +213 -214 u=11 imp:p=1.0E+00 $ void around fuel 2031 6 -7.940 (-211:+212:-213:+214) &

Page 5-38

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

+271 -272 +273 -274 u=11 imp:p=1.0E+00 $ steel liner 2041 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=11 imp:p=1.0E+00 $ aluminum sandwich 2061 0 +295 -296 +297 -298 -205 u=11 imp:p=1.0E+00 $ void gap 2051 4 -2.595 +295 -296 +297 -298 +205 -208 u=11 imp:p=1.0E+00 $ bottom end fitting 2071 1 -3.949 +295 -296 +297 -298 +208 -250 u=11 imp:p=1.0E+00 $ fuel 2081 2 -1.543 +295 -296 +297 -298 +250 -255 u=11 imp:p=1.0E+00 $ plenum 2091 3 -1.970 +295 -296 +297 -298 +255 -260 u=11 imp:p=1.0E+00 $ top end fitting 2101 0 +295 -296 +297 -298 +260 u=11 imp:p=1.0E+00 $ void gap c

c periphery fuel assys 2012 10 -2.45 (-275:+276:-277:+278) u=12 imp:p=1.0E+00 $ boral gap 2022 0 (-295:+296:-297:+298) &

+211 -212 +213 -214 u=12 imp:p=1.0E+00 $ void around fuel 2032 6 -7.940 (-211:+212:-213:+214) &

+271 -272 +273 -274 u=12 imp:p=1.0E+00 $ steel liner 2042 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=12 imp:p=1.0E+00 $ aluminum sandwich 2052 4 -2.595 +295 -296 +297 -298 -208 u=12 imp:p=1.0E+00 $ bottom end fitting 2062 0 +295 -296 +297 -298 -205 u=12 imp:p=1.0E+00 $ void gap 2072 1 -3.949 +295 -296 +297 -298 +208 -250 u=12 imp:p=1.0E+00 $ fuel 2082 2 -1.543 +295 -296 +297 -298 +250 -255 u=12 imp:p=1.0E+00 $ plenum 2092 3 -1.970 +295 -296 +297 -298 +255 -260 u=12 imp:p=1.0E+00 $ top end fitting 2102 0 +295 -296 +297 -298 +260 u=12 imp:p=1.0E+00 $ void gap c

c c empty compartment and periphery basket 2013 10 -2.702 (-275:+276:-277:+278) u=2 imp:p=1.0E+00 $ boral gap 2043 10 -2.702 (-271:+272:-273:+274) &

+275 -276 +277 -278 u=2 imp:p=1.0E+00 $ aluminum sandwich 2073 0 +271 -272 +273 -274 u=2 imp:p=1.0E+00 $ empty basket 241 10 -2.702 +272:-273 u=5 imp:p=1.0E+00 $ solid aluminum 245 0 -272 +273 u=5 imp:p=1.0E+00 $ empty assy 251 10 -2.702 -271:-273 u=6 imp:p=1.0E+00 $ basket 255 0 +271 +273 u=6 imp:p=1.0E+00 $ empty assy 261 10 -2.702 -271:+274 u=7 imp:p=1.0E+00 $ basket 265 0 +271 -274 u=7 imp:p=1.0E+00 $ empty assy 271 10 -2.702 +272:+274 u=8 imp:p=1.0E+00 $ basket 275 0 -272 -274 u=8 imp:p=1.0E+00 $ empty assy 299 0 -292 +291 -294 +293 u=9 imp:p=1.0E+00 &

lat=1 fill=-1:8 -5:5 0:0 c

c x coords c -1 0 1 2 3 4 5 6 7 8 c

9 9 9 9 9 9 9 9 9 9 $ -5 9 9 9 2 2 2 2 9 9 9 $ -4 9 7 8 12 12 12 12 7 8 9 $ -3 9 2 12 11 11 11 11 12 2 9 $ -2 9 2 12 11 11 11 11 12 2 9 $ -1 9 12 11 11 11 11 11 11 12 9 $ 0 9 2 12 11 11 11 11 12 2 9 $ 1 9 2 12 11 11 11 11 12 2 9 $ 2 Page 5-39

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 9 6 5 12 12 12 12 6 5 9 $ 3 9 9 9 2 2 2 2 9 9 9 $ 4 9 9 9 9 9 9 9 9 9 9 $ 5 c

c cavity cells 23 0 +201 -261 -2990 fill=9 imp:p=1.0E+00 24 6 -7.940 +201 -261 +2990 -299 imp:p=1.0E+00 $ outer SST inserts 2771 10 -2.702 +201 -261 +299 -282 -2991 imp:p=1.0E+00 $ these are the rails 2772 10 -2.702 +201 -261 +299 -282 +2991 -2992 imp:p=1.0E+00 2773 10 -2.702 +201 -261 +299 -282 +2992 -2993 imp:p=1.0E+00 2774 10 -2.702 +201 -261 +299 -282 +2993 -2994 imp:p=1.0E+00 2775 10 -2.702 +201 -261 +299 -282 +2994 -2995 imp:p=1.0E+00 2776 10 -2.702 +201 -261 +299 -282 +2995 -2996 imp:p=1.0E+00 2778 10 -2.702 +201 -261 +299 -282 +2996 imp:p=1.0E+00 c

c main shield shell cells 301 5 -7.702 (261 -511 -801):&

(802 801 282 -311):(419 -201 -802) imp:p=1.0E+00 302 5 -7.702 (511 -512 -801):&

(802 801 311 -312):(418 -419 -802) imp:p=1.0E+00 303 5 -7.702 (512 -513 -801):&

(802 801 312 -313):(417 -418 -802) imp:p=1.0E+00 304 5 -7.702 (513 -514 -801):&

(802 801 313 -314):(416 -417 -802) imp:p=1.0E+00 305 5 -7.702 (514 -515 -801):&

(802 801 314 -315):(415 -416 -802) imp:p=1.0E+00 306 5 -7.702 (515 -516 -801):&

(802 -516 315 -316):(414 -415 -802) imp:p=1.0E+00 307 5 -7.702 (802 -516 316 -317):(413 -414 -802) imp:p=1.0E+00 308 5 -7.702 (802 -516 317 -318):(412 -413 -802) imp:p=1.0E+00 309 5 -7.702 (802 -516 318 -319):(411 -412 -802) imp:p=1.0E+00 310 5 -7.702 (802 -516 319 -320):(410 -411 -802) imp:p=1.0E+00 c

c top of lid 501 5 -7.702 +516 -517 -521 imp:p=1.0E+00 502 5 -7.702 +517 -518 -521 imp:p=1.0E+00 503 5 -7.702 +518 -519 -521 imp:p=1.0E+00 504 5 -7.702 +519 -520 -521 imp:p=3.0E+00 505 10 -2.702 +520 -522 -320 imp:p=3.0E+00 $ Al impact limiter spacer on top c

c top impact limiters 601 0 (+622 -750 +320 -651) imp:p=1.0E+00 602 0 (+622 -750 +651 -672) imp:p=1.0E+00 603 8 -.125 (+750 -752 +320 -651) imp:p=1.0E+00 604 8 -.125 (+750 -752 +651 -672) imp:p=1.0E+00 605 10 -2.702 (+516 -518 +521 -318) imp:p=1.0E+00 606 10 -2.702 (+518 -520 +521 -318):&

(+516 -520 +318 -320) imp:p=3.0E+00 607 9 -.387 (+522 -721 -651):&

(+752 -522 +320 -651) imp:p=3.9E+00 608 9 -.387 (+752 -721 +651 -672):&

(+721 -722 -672) imp:p=4.2E+00 609 9 -.387 +722 -723 -672 imp:p=4.4E+00 610 9 -.387 +723 -724 -672 imp:p=4.1E+00 611 9 -.387 +724 -725 -672 imp:p=4.5E+00 612 9 -.387 +725 -726 -672 imp:p=5.1E+00 613 9 -.387 +726 -727 -672 imp:p=6.5E+00 614 9 -.387 +727 -728 -672 imp:p=8.4E+00 615 9 -.387 +728 -729 -672 imp:p=1.0E+01 616 9 -.387 +729 -730 -672 imp:p=1.3E+01 617 9 -.387 +730 -731 -672 imp:p=1.5E+01 618 9 -.387 +731 -732 -672 imp:p=2.0E+01 Page 5-40

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 619 9 -.387 +732 -733 -672 imp:p=2.8E+01 620 9 -.387 +733 -734 -672 imp:p=4.0E+01 621 9 -.387 +734 -735 -672 imp:p=6.5E+01 622 8 -.125 +735 -762 -672 imp:p=1.0E+02 623 6 -7.940 +762 -736 -672 imp:p=1.0E+02 c

632 0 +622 -750 +651 +672 -810 imp:p=1.1E+01 634 8 -.125 +750 -752 +651 +672 -810 imp:p=6.3E+00 638 9 -.387 +752 -723 +651 +672 -810 imp:p=5.5E+00 640 9 -.387 +723 -725 +672 -810 imp:p=6.2E+00 642 9 -.387 +725 -727 +672 -810 imp:p=9.4E+00 644 9 -.387 +727 -729 +672 -810 imp:p=1.4E+01 646 9 -.387 +729 -731 +672 -810 imp:p=2.2E+01 648 9 -.387 +731 -733 +672 -810 imp:p=4.0E+01 650 9 -.387 +733 -735 +672 -810 imp:p=9.1E+01 652 8 -.125 +735 -762 +672 -810 imp:p=2.4E+02 653 6 -7.940 +762 -736 +672 -810 imp:p=2.4E+02 c

662 0 +622 -750 +651 +810 -811 imp:p=2.4E+01 664 8 -.125 +750 -752 +651 +810 -811 imp:p=1.5E+01 668 9 -.387 +752 -723 +651 +810 -811 imp:p=1.2E+01 672 9 -.387 +723 -727 +810 -811 imp:p=1.7E+01 676 9 -.387 +727 -731 +810 -811 imp:p=3.2E+01 680 9 -.387 +731 -735 +810 -811 imp:p=1.2E+02 682 8 -.125 +735 -762 +810 -811 imp:p=6.2E+02 683 6 -7.940 +762 -736 +810 -811 imp:p=6.2E+02 c

c side neutron shield 851 7 -1.642 +905 -904 +650 -651 +861 imp:p=1.0E+00 852 7 -1.642 +905 -904 +651 -670 +861 imp:p=1.0E+00 853 5 -7.702 +621 -622 +320 -671 -906 907 imp:p=1.0E+00 854 5 -7.702 +601 -602 +320 -671 -906 907 imp:p=1.0E+00 855 5 -7.702 +905 -904 +671 -672 +861 imp:p=1.0E+00 856 7 -1.642 +905 -904 +320 -650 +861 imp:p=1.0E+00 857 7 -1.642 +905 -904 +670 -671 +861 imp:p=1.0E+00 c Neutron shielding - Flat - Top 858 7 -1.642 +900 -621 +650 -651 +861 -906 907 imp:p=1.0E+00 859 7 -1.642 +900 -621 +651 -670 +861 -906 907 imp:p=1.0E+00 860 7 -1.642 +900 -621 +320 -650 +861 -906 907 imp:p=1.0E+00 861 7 -1.642 +900 -621 +670 -671 +861 -906 907 imp:p=1.0E+00 862 7 -1.642 +904 -900 +650 -651 +861 -906 907 imp:p=1.0E+00 863 7 -1.642 +904 -900 +651 -670 +861 -906 907 imp:p=1.0E+00 864 7 -1.642 +904 -900 +320 -650 +861 -906 907 imp:p=1.0E+00 865 7 -1.642 +904 -900 +670 -671 +861 -906 907 imp:p=1.0E+00 c Neutron shielding - Flat - Bottom 866 7 -1.642 +602 -901 +650 -651 +861 -906 907 imp:p=1.0E+00 867 7 -1.642 +602 -901 +651 -670 +861 -906 907 imp:p=1.0E+00 868 7 -1.642 +602 -901 +320 -650 +861 -906 907 imp:p=1.0E+00 869 7 -1.642 +602 -901 +670 -671 +861 -906 907 imp:p=1.0E+00 870 7 -1.642 +901 -905 +650 -651 +861 -906 907 imp:p=1.0E+00 871 7 -1.642 +901 -905 +651 -670 +861 -906 907 imp:p=1.0E+00 872 7 -1.642 +901 -905 +320 -650 +861 -906 907 imp:p=1.0E+00 873 7 -1.642 +901 -905 +670 -671 +861 -906 907 imp:p=1.0E+00 c

874 5 -7.702 +900 -622 +671 -672 +861 -906 907 imp:p=1.0E+00 875 5 -7.702 +904 -900 -907 -672 +861 imp:p=1.0E+00 876 5 -7.702 +900 +861 +903 -907 -622 -672 imp:p=1.0E+00 877 5 -7.702 +900 +861 +906 -902 -622 -672 imp:p=1.0E+00 878 5 -7.702 +904 -900 +906 -672 +861 imp:p=1.0E+00 879 5 -7.702 +904 -900 +907 -906 -672 +671 imp:p=1.0E+00 880 5 -7.702 +601 -901 +671 -672 +861 -906 907 imp:p=1.0E+00 881 5 -7.702 +901 -905 -907 -672 +861 imp:p=1.0E+00 882 5 -7.702 +901 -905 +906 -672 +861 imp:p=1.0E+00 Page 5-41

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 883 5 -7.702 +901 -905 +907 -906 -672 +671 imp:p=1.0E+00 884 5 -7.702 +601 +864 +903 -907 -901 -672 imp:p=1.0E+00 885 5 -7.702 +601 +864 +906 -902 -901 -672 imp:p=1.0E+00 c Air 886 0 +900 -622 +861 -903 -672 imp:p=1.0E+00 887 0 +900 -622 +861 +902 -672 -908 imp:p=1.0E+00 888 0 +900 -622 +861 +902 -672 +908 imp:p=1.0E+00 889 0 +601 -901 -903 -672 imp:p=1.0E+00 890 0 +601 -901 +902 -672 imp:p=1.0E+00 c bottom impact limiters c

701 8 -.125 (+751 -601 +320 -651 -875):&

(+751 -601 +320 -651 876) imp:p=1.0E+00 702 8 -.125 (+751 -601 +651 -672 -875):&

(+751 -601 +651 -672 876) imp:p=1.0E+00 7031 9 -.387 (+410 -751 +320 -651 ) &

(-875:876:-879) imp:p=1.0E+00 7032 9 -.387 (+716 -410 -651) imp:p=2.3E+00 7041 9 -.387 (+716 -751 +651 -672) &

(-875:876:-879) imp:p=1.0E+00 7042 9 -.387 (+715 -716 -672) imp:p=2.5E+00 705 9 -.387 +714 -715 -672 imp:p=2.5E+00 706 9 -.387 +713 -714 -672 imp:p=2.8E+00 707 9 -.387 +712 -713 -672 imp:p=3.2E+00 708 9 -.387 +711 -712 -672 imp:p=3.4E+00 709 9 -.387 +710 -711 -672 imp:p=4.0E+00 710 9 -.387 +709 -710 -672 imp:p=5.0E+00 711 9 -.387 +708 -709 -672 imp:p=6.5E+00 712 9 -.387 +707 -708 -672 imp:p=8.0E+00 713 9 -.387 +706 -707 -672 imp:p=9.8E+00 714 9 -.387 +705 -706 -672 imp:p=1.4E+01 715 9 -.387 +704 -705 -672 imp:p=1.9E+01 716 9 -.387 +703 -704 -672 imp:p=2.9E+01 717 9 -.387 +761 -703 -672 imp:p=2.9E+01 718 8 -.125 +702 -761 -672 imp:p=6.7E+01 719 6 -7.940 +701 -702 -672 imp:p=6.7E+01 c

721 0 +753 -601 +672 -810 &

(-875:876:-877:878) imp:p=5.4E+00 722 8 -.125 +751 -753 +672 -810 &

(-875:876:-877:878) imp:p=5.4E+00 724 9 -.387 +715 -751 +672 -810 &

(-875:876:-877:878:-879) imp:p=4.3E+00 726 9 -.387 +713 -715 +672 -810 imp:p=4.5E+00 728 9 -.387 +711 -713 +672 -810 imp:p=6.1E+00 730 9 -.387 +709 -711 +672 -810 imp:p=9.5E+00 732 9 -.387 +707 -709 +672 -810 imp:p=1.6E+01 734 9 -.387 +705 -707 +672 -810 imp:p=2.3E+01 736 9 -.387 +761 -705 +672 -810 imp:p=5.2E+01 738 8 -.125 +702 -761 +672 -810 imp:p=1.9E+02 739 6 -7.940 +701 -702 +672 -810 imp:p=1.9E+02 c

741 0 +753 -601 +810 -811 imp:p=1.4E+01 742 8 -.125 +751 -753 +810 -811 imp:p=1.4E+01 744 9 -.387 +715 -751 +810 -811 imp:p=1.1E+01 748 9 -.387 +711 -715 +810 -811 imp:p=1.3E+01 752 9 -.387 +707 -711 +810 -811 imp:p=2.4E+01 756 9 -.387 +761 -707 +810 -811 imp:p=8.0E+01 758 8 -.125 +702 -761 +810 -811 imp:p=5.4E+02 759 6 -7.940 +701 -702 +810 -811 imp:p=5.4E+02 c

760 0 (864 867 875 -876 320 -878 879 -601):&

(864 -867 875 -876 320 +877 879 -601) imp:p=1.0E+00 Page 5-42

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c

c tally cells 900 0 +601 -622 +672 -899 +861 imp:p=4.0E+00 901 0 +601 -622 +899 -810 +861 imp:p=4.0E+00 902 0 +601 -622 +810 -811 imp:p=1.2E+01 903 11 -1.225E-3 +701 -736 +811 -812 imp:p=1.2E+02 $ comment these out for importance optimization 904 11 -1.225E-3 (+701 -736 +812 -814):&

(+736 -815 -814):(+816 -701 -814) imp:p=6.2E+02 905 11 -1.225E-3 (+816 -815 +814 -813):&

(+815 -817 -813):(+818 -816 -813) imp:p=6.2E+02 906 11 -1.225E-3 (+11 13) (860:+817:-818) &

(+818:-11:810) (-817:12:810) imp:p=6.2E+02 907 11 -1.225E-3 +851 -818 -810 imp:p=6.2E+02 $ far cells at end of rail car 908 11 -1.225E-3 +850 -851 -810 imp:p=6.2E+02 909 11 -1.225E-3 +11 -850 -810 imp:p=6.2E+02 910 11 -1.225E-3 +817 -852 -810 imp:p=6.2E+02 911 11 -1.225E-3 +852 -853 -810 imp:p=6.2E+02 912 11 -1.225E-3 +853 810 imp:p=6.2E+02 913 11 -1.225E-3 (+11 860) (813:+817:-818) &

(+818:-11:810) (-817:12:810) imp:p=6.2E+02 c

c upper trunnions 801 5 -7.702 (-867 +320 +870 -861):&

(+867 +320 -871 -861) imp:p=1.0E+00 802 5 -7.702 (+869 -870 +863 -861):&

(+871 -872 +863 -861) imp:p=1.0E+00 803 5 -7.702 (+868 -869 +863 -862):&

(+872 -873 +863 -862) imp:p=1.0E+00 804 7 -1.642 (+868 -870 -863):&

(+871 -873 -863) imp:p=1.0E+00 805 0 (-869 -810 -861 (+862:-868)):&

(+872 -810 -861 (+862:+873)) imp:p=1.0E+00 c

c lower trunnions 811 5 -7.702 (-867 +320 +911 -864):&

(+867 +320 -912 -864) imp:p=4.0E+00 812 5 -7.702 (+869 -911 +866 -864):&

(+912 -872 +866 -864) imp:p=2.5E+00 813 5 -7.702 (+868 -869 +866 -865):&

(+872 -873 +866 -865) imp:p=5.0E+00 814 0 (+868 -911 -866):&

(+912 -873 -866) imp:p=5.4E+00 815 0 (+877 -869 +865 -864):&

(+872 -878 +865 -864) imp:p=5.0E+00 816 0 ( +877 -868 -865):&

(+873 -878 -865) imp:p=5.4E+00 c ==============================================================================

c = =

c = SURFACE CARDS =

c = =

c ==============================================================================

c c model boundaries c initial model had rail car at 41' 8" long c rail cars typically are 40 to 60 feet c PacTec typical transport car (for OCRWM) 49' 6" c set outer bounds at 60 feet c assume cask CG is middle of active fuel (within 6 inches of actual) 11 pz -914.4 $ end of rail car 12 pz 914.4 $ end of rail car Page 5-43

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 13 cz 652.4 $ 5 meter from (vertical planes) side of rail car c

c fuel/basket region 201 pz -190.70 $ bottom of lower end fitting (bottom cavity) 205 pz -190.7032 $ top of lower end fitting (3.08")

208 pz -182.88 $ bottom of active fuel (144")

211 px -88.7034 $ steel liner inner (8.05 inch typ. square) 212 px -68.20565 $ steel liner modeled as 13 ga. 0.09" thick not 0.10" 213 py -10.2494 $

214 py 10.2494 $

250 pz 182.88 $ top of active fuel 255 pz 201.02068 $ top of plenum (7.142")

260 pz 209.91 $ top of upper end fitting (3.5")

261 pz 223.32 $ top of cavity (void) 271 px -88.93175 $ steel liner outer (8.05 + .2 inch square) 272 px -67.97675 $

273 py -10.4775 $

274 py 10.4775 $

275 px -89.56675 $ boral gap (8.05 + .7 inch square) 276 px -67.34175 $

277 py -11.1125 $

278 py 11.1125 $

282 cz 91.44 $ inner radius of cavity (72 inch diameter) 291 px -89.662 $ basket lattice unit cell (8.05+.775 inch square) 292 px -67.2465 $ origin (center of assy) @ x=-78.45425 y=0 293 py -11.20775 $

294 py 11.20775 $

295 px -88.31326 $ fuel envelope 7.763 x 7.763 inch 296 px -68.59524 $

297 py -9.85901 $

298 py 9.85901 $

299 cz 90.932 $ assumes .2" instead of 3/8" outer basket ("rails") to conserve mass 2990 cz 90.867 $ assumes 0.026" for steel inserts on periphery 2991 py -79.4385 $ rail cutting surfaces 2992 py -55.0545 $

2993 py -12.192 $

2994 py 12.192 $

2995 py 55.0545 $

2996 py 79.4385 $

c c wall gamma shield 311 cz 93.853 $ split primary gamma shield into 10 layers 312 cz 96.266 $ layer thickness = 2.413 cm or 0.95 inch 313 cz 98.679 $ also applies to bottom and top (diff. thickness) 314 cz 101.092 $

315 cz 103.505 $

316 cz 105.918 $

317 cz 108.331 $

318 cz 110.744 $

319 cz 113.157 $

320 cz 115.57 $ outside cask body c

c bottom gamma shield 410 pz -216.73 $ cask bottom 411 pz -214.127 $ layer thickness = 2.603 cm or 1.025 inch 412 pz -211.524 $

413 pz -208.921 $

414 pz -206.318 $

415 pz -203.715 $

416 pz -201.112 $

417 pz -198.509 $

418 pz -195.906 $

Page 5-44

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 419 pz -193.303 $

c c top gamma shield 511 pz 225.987 $ layer thickness = 2.667 cm or 1.05 inch 512 pz 228.654 $

513 pz 231.321 $

514 pz 233.988 $

515 pz 236.655 $

516 pz 238.56 $ top of cask walls 517 pz 241.989 $

518 pz 244.656 $

519 pz 247.323 $

520 pz 249.99 $

521 cz 105.09 $ outer radius of top part of lid 522 pz 252.022 $ assume 0.8" thick to account for tolerences and bolt holes c

c neutron shields c 601 pz -186.25 $ bottom edge of side resin lower housing 601 pz -185.615 $ bottom edge of side resin lower housing c 602 pz -184.34 $ top edge of side resin lower housing 602 pz -183.705 $ top edge of side resin lower housing 621 pz 203.00 $ bottom edge of side resin upper housing 622 pz 204.91 $ top edge of side resin upper housing 650 cz 116.1542 $ homogenized inner aluminum box 651 cz 121.285 $ splitting shell for neutron shielding 670 cz 126.4158 $ homogenized outer aluminum box 671 cz 127.0 $ outer radius of resin shield on side 672 cz 128.27 $ outer radius of resin housing on side c

c impact limiters 701 pz -311.98 702 pz -311.345 703 pz -299.05 704 pz -293.17 705 pz -287.29 706 pz -281.41 707 pz -275.53 708 pz -269.65 709 pz -263.77 710 pz -257.89 711 pz -252.01 712 pz -246.13 713 pz -240.25 714 pz -234.37 715 pz -228.49 716 pz -222.61 c

721 pz 255.87 722 pz 261.75 723 pz 267.63 724 pz 273.51 725 pz 279.39 726 pz 285.27 727 pz 291.15 728 pz 297.03 729 pz 302.91 730 pz 308.79 731 pz 314.67 732 pz 320.55 733 pz 326.43 734 pz 332.31 735 pz 338.255 Page 5-45

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 736 pz 345.24 c

750 pz 220.78 $ bottom of top impact limiter 751 pz -195.14 $ balsa/redwood boundary 752 pz 228.4 $ balsa/redwood boundary 753 pz -187.52 $ balsa/redwood boundary 761 pz -304.995 $ balsa/redwood boundary 762 pz 344.605 $ balsa/redwood boundary c

c miscellaneous surfaces 801 kz 122.2547369 0.8185941 +1 $ upper splitting mating plane 802 kz -92.06 0.8593425329 -1 $ lower splitting mating plane 899 cz 128.271 $ void cell outside cask 810 cz 152.4 $ vertical planes on side (10' wide rail car) 811 cz 182.88 $ top/bottom @ 12' diameter impact limiters 812 cz 198.12 $ top/bottom @ 13' diameter impact limiters (not used) 813 cz 352.4 $ 2 meter from vertical planes on side 860 cz 383.0 $ 2 meter from vertical impact limiter 814 cz 215.57 $ 1 meter from cask body (HAC) 815 pz 349.99 $ 1 meter from cask top lid (HAC) 816 pz -316.73 $ 1 meter from cask bottom (HAC) 817 pz 545.24 $ 2 meter from top impact limiter 818 pz -511.98 $ 2 meter from bottom impact limiter 824 pz -190 $ axial tally segments 825 pz -170 $ 20 cm wide 826 pz -150 $ centered on fuel midplane 827 pz -130 828 pz -110 829 pz -90 830 pz -70 831 pz -50 832 pz -30 833 pz -10 834 pz 10 835 pz 30 836 pz 50 837 pz 70 838 pz 90 839 pz 110 840 pz 130 841 pz 150 842 pz 170 843 pz 190 844 pz 210 845 pz 230 846 pz 250 c occupied position surfaces 850 pz -762.0 $ 40 foot rail car 851 pz -609.6 $ 50 foot rail car 852 pz 609.6 853 pz 762.0 882 cz 30 $ radial segmentation 883 cz 55 885 cz 140.57 886 cz 165.57 887 cz 232.88 888 cz 282.88 861 c/y 0 177.6 15.254 $ top trunnion base radius 862 c/y 0 177.6 14.2875 $ top trunnion 863 c/y 0 177.6 6.35 $ top trunnion resin radius 864 c/y 0 -197.685 12.0624 $ bottom trunnion base (9.498" dia.)

865 c/y 0 -197.685 11.2776 $ bottom trunnion 866 c/y 0 -197.685 5.715 $ bottom trunnion hole radius Page 5-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 867 py 0 $ trunnion ambiguious surface 868 py -132.715 $ NEW distance between trunnion head 104.5" 869 py -128.27 870 py -123.825 871 py 123.825 872 py 128.27 873 py 132.715 $ NEW distance between trunnion head 104.5" c

c lower trunnion "channel" in impact limiter 875 px -12.6974 $ these surfs form a box around lower trunnions 876 px 12.6974 $ 1/4" gap is assumed 877 py -136.8552 $

878 py 136.8552 $

879 pz -210.3824 $

c 900 pz 133.785 $ 17.25" from surf. center of top trunnion 901 pz -153.865 $ estimate 17.25" from surf. center of trunnion 902 py 124.7648 $ distance between surf. 902-903 flat -> 98.24" 903 py -124.7648 $

904 pz 132.515 905 pz -152.595 906 py 123.4948 $ 0.5" from surf. 902 907 py -123.4948 $ 0.5" from surf. 903 c

908 pz 177.6 $

c 911 py -122.047 912 py 122.047 c ==============================================================================

c = =

c = DATA CARDS =

c = =

c ==============================================================================

c c materials c

c **************************************************************

c In-Core Region c Density = 3.949 g/cm^3; Composition by weight fraction c 3.0 w/o enrichment c 410 kg U c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m1 92234 -0.000195 $ U-234 92235 -0.021902 $ U-235 92236 -0.000101 $ U-236 92238 -0.707883 $ U-238 8016 -0.098148 $ O 6012 -0.000011 $ C 14000 -0.000377 $ Si 15031 -0.000006 $ P 24000 -0.004200 $ Cr 22000 -0.000239 $ Ti 25055 -0.000276 $ Mn 26000 -0.010400 $ Fe 28000 -0.008289 $ Ni 40000 -0.145805 $ Zr 50000 -0.002152 $ Sn 72000 -0.000015 $ Hf c 1001 0.000000 $ H c

Page 5-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c **************************************************************

c Plenum c Density = 1.543 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m2 6012 -0.00045 $ C 14000 -0.00714 $ Si 15031 -0.00025 $ P 24000 -0.11569 $ Cr 22000 -0.00156 $ Ti 25055 -0.01115 $ Mn 26000 -0.38641 $ Fe 28000 -0.09859 $ Ni 40000 -0.37321 $ Zr 50000 -0.00551 $ Sn 72000 -0.00004 $ Hf c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Top Region c Density = 1.970 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m3 6012 -0.00074 $ C 14000 -0.01112 $ Si 15031 -0.00042 $ P 24000 -0.18702 $ Cr 22000 -0.00187 $ Ti 25055 -0.01851 $ Mn 26000 -0.63795 $ Fe 28000 -0.14238 $ Ni c 40000 0.00000 $ Zr c 50000 0.00000 $ Sn c 72000 0.00000 $ Hf c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Bottom Region c Density = 2.595 g/cm^3; Composition by weight fraction c Chemical composition from SCALE Standard Comp. Library c **************************************************************

c m4 6012 -0.00080 $ C 14000 -0.01000 $ Si 15031 -0.00045 $ P 24000 -0.19000 $ Cr c 22000 0.00000 $ Ti 25055 -0.02000 $ Mn 26000 -0.68375 $ Fe 28000 -0.09500 $ Ni c 40000 0.00000 $ Zr c 50000 0.00000 $ Sn c 72000 0.00000 $ Hf c 8016 0.00000 $ O c 1001 0.00000 $ H c

c **************************************************************

c Carbon Steel c Density = 7.8212 g/cm^3 SCALE Standard Comp. Library Page 5-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c **************************************************************

c m5 6012 -0.0100 $ C 26000 -0.9900 $ Fe c

c **************************************************************

c Stainless Steel 304 c Density = 7.94 g/cm^3 SCALE Standard Comp. Library c **************************************************************

c m6 26000 -0.68375 $ Fe 24000 -0.19000 $ Cr 28000 -0.09500 $ Ni 25055 -0.02000 $ Mn 14000 -0.01000 $ Si 6012 -0.00080 $ C 15031 -0.00045 $ P c

c **************************************************************

c Homogenized Neutron Resin/Aluminum Shield c

Reference:

Page 6, TN Calc. 1042-08, Rev. 0 c Based on TN-24 Resin c Density = 1.69 g/cm^3 c B-10 and B-11 based on natural abundance c Note Al is in resin and the aluminum boxes c **************************************************************

c m7 1001 -0.0418 $ H-1 5010 -0.0016 $ B-10 5011 -0.0071 $ B-11 6000 -0.2908 $ C 8016 -0.3455 $ O 13027 -0.2958 $ Al 30000 -0.0175 $ Zn c

c **************************************************************

c Balsa Wood c TN-68 SAR c Density = 0.125 g/cm^3 c **************************************************************

c m8 6012 0.2857 $ C 8016 0.2381 $ O 1001 0.4762 $ H c

c **************************************************************

c Redwood c Assume same compositin as Balsa c TN-68 SAR c Density = 0.387 g/cm^3 c **************************************************************

c m9 6012 0.2857 $ C 8016 0.2381 $ O 1001 0.4762 $ H c

c **************************************************************

c Pure Aluminum c Density = 2.702 g/cm^3 SCALE Standard Comp. Library c **************************************************************

c m10 13027 -1.0 $ Al c

Page 5-49

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c **************************************************************

c AIR: ANSI/ANS-6.6.1 Dry air c Density = 0.001225 g/cm^3 c Composition by weight fraction c **************************************************************

c m11 6012 -0.00014 $ C 7014 -0.75519 $ N 8016 -0.23179 $ O 18000 -0.01288 $ Ar c

c **************************************************************

c BORAL: TN40 STORAGE CRITICALITY SPECIFICATION c Density = 2.45 g/cm^3 (770 lbs of boral, about 95.8% theo.)

c Composition by weight fraction c Based on Calculation 1042-6 (TN40 Storage Crit) c 0.01 gm/cm^2 B-10 areal density c core thickness of 0.025" c panel thickness of 0.075" c redistribution of wt. frac. to simulate 75% B4C c **************************************************************

c actual wt. frac m12 6012 -0.066 $ 0.088 13027 -0.695 $ 0.594 5010 -0.044 $ 0.058 5011 -0.195 $ 0.060 c

c source c

sdef cel=d1 x=d2 y=d3 z=fcel=d4 erg=fcel=d5 c

c sample cells according to strength c there are 22 inner and 18 outer

si1 sp1 c 39GWD 39GWD l d $ 17.5yr 16 year 23:299:2051 1.83E+13 $ 3.99E+13 4.86E+13 inner assys lower end fitting 23:299:2071 3.92E+16 $ 4.49E+16 4.72E+16 inner assys fuel 23:299:2081 6.16E+13 $ 8.94E+13 1.27E+14 inner assys plenum 23:299:2091 3.65E+13 $ 4.92E+13 7.22E+13 inner assys upper end fitting 23:299:2052 1.50E+13 $ 3.27E+13 3.98E+13 outer assys lower end fitting 23:299:2072 3.21E+16 $ 3.67E+16 3.86E+16 outer assys fuel 23:299:2082 5.04E+13 $ 7.31E+13 1.04E+14 outer assys plenum 23:299:2092 2.99E+13 $ 4.03E+13 5.90E+13 outer assys upper end fitting c

c sample source/fuel cell c sample volume uniformly c use lattice element (0,0) c c X DIMENSIONS Y DIMENSIONS

si2 sp2 si3 sp3

-88.31326 0 -9.85901 0

-68.59524 1 9.85901 1 c

c zone dependent axial distirubtions c burnup is taken from Prairie site specific calculation c cask center c

ds4 s 41 44 42 43 $ 41-43 are end fittings/plenum 41 45 42 43 $ 44 and 45 are fuel zones c

c lower end fit plenum upper end fit

si41 sp41 si42 sp42 si43 sp43 Page 5-50

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

-190.7032 0 182.88 0 201.02068 0

-182.88 1 201.02068 1 223.32 1 c

c INNER ASSYS OUTER ASSYS

si44 sp44 si45 sp45

-182.88 0 -182.88 0

-152.40 0.0556 -152.40 0.0556

-121.92 0.0818 -121.92 0.0818

-91.44 0.0889 -91.44 0.0889

-60.96 0.0921 -60.96 0.0921

-30.48 0.0921 -30.48 0.0921 0.00 0.0921 0.00 0.0921 30.48 0.0920 30.48 0.0920 60.96 0.0913 60.96 0.0913 91.44 0.0909 91.44 0.0909 121.92 0.0889 121.92 0.0889 152.40 0.0786 152.40 0.0786 182.88 0.0556 182.88 0.0556 c

c zone dependent energy distributions c

ds5 s 51 52 53 54 $ each cell has independent energy distribution 55 56 57 58 $

c c INNER ASSYS GAMMA SPECTRA c

c lower end fuel plenum upper end

si51 sp51 si52 sp52 si53 sp53 si54 sp54 0.01 0.00E+00 0.01 0.00E+00 0.01 0.00E+00 0.01 0.00E+00 0.05 1.15E+10 0.05 4.82E+14 0.05 1.58E+10 0.05 7.03E+09 0.1 2.19E+09 0.1 1.41E+14 0.1 2.81E+09 0.1 1.29E+09 0.2 5.27E+08 0.2 8.83E+13 0.2 7.15E+08 0.2 3.11E+08 0.3 2.62E+07 0.3 2.70E+13 0.3 3.61E+07 0.3 1.55E+07 0.4 3.43E+07 0.4 1.81E+13 0.4 5.27E+07 0.4 2.02E+07 0.6 2.17E+06 0.6 1.47E+13 0.6 1.94E+08 0.6 1.28E+06 0.8 7.54E+05 0.8 9.70E+14 0.8 8.56E+08 0.8 2.83E+08 1 2.84E+07 1 8.23E+12 1 7.65E+08 1 2.89E+08 1.33 6.38E+11 1.33 2.75E+13 1.33 2.17E+12 1.33 1.29E+12 1.66 1.80E+11 1.66 5.38E+12 1.66 6.11E+11 1.66 3.63E+11 2 1.64E-06 2 4.63E+10 2 1.47E+01 2 1.10E-03 2.5 4.27E+06 2.5 2.48E+09 2.5 5.48E+06 2.5 2.51E+06 3 6.63E+03 3 1.73E+08 3 8.50E+03 3 3.90E+03 4 6.28E-15 4 2.68E+07 4 1.52E-11 4 5.62E-12 5 0.00E+00 5 9.04E+06 5 0.00E+00 5 0.00E+00 6.5 0.00E+00 6.5 3.63E+06 6.5 0.00E+00 6.5 0.00E+00 8 0.00E+00 8 7.12E+05 8 0.00E+00 8 0.00E+00 10 0.00E+00 10 1.51E+05 10 0.00E+00 10 0.00E+00 c

c OUTER ASSYS GAMMA SPECTRA c

c lower end fuel plenum upper end

si55 sp55 si56 sp56 si57 sp57 si58 sp58 0.01 0.00E+00 0.01 0.00E+00 0.01 0.00E+00 0.01 0.00E+00 0.05 1.15E+10 0.05 4.82E+14 0.05 1.58E+10 0.05 7.03E+09 0.1 2.19E+09 0.1 1.41E+14 0.1 2.81E+09 0.1 1.29E+09 0.2 5.27E+08 0.2 8.83E+13 0.2 7.15E+08 0.2 3.11E+08 0.3 2.62E+07 0.3 2.70E+13 0.3 3.61E+07 0.3 1.55E+07 0.4 3.43E+07 0.4 1.81E+13 0.4 5.27E+07 0.4 2.02E+07 0.6 2.17E+06 0.6 1.47E+13 0.6 1.94E+08 0.6 1.28E+06 0.8 7.54E+05 0.8 9.70E+14 0.8 8.56E+08 0.8 2.83E+08 1 2.84E+07 1 8.23E+12 1 7.65E+08 1 2.89E+08 1.33 6.38E+11 1.33 2.75E+13 1.33 2.17E+12 1.33 1.29E+12 1.66 1.80E+11 1.66 5.38E+12 1.66 6.11E+11 1.66 3.63E+11 Page 5-51

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2 1.64E-06 2 4.63E+10 2 1.47E+01 2 1.10E-03 2.5 4.27E+06 2.5 2.48E+09 2.5 5.48E+06 2.5 2.51E+06 3 6.63E+03 3 1.73E+08 3 8.50E+03 3 3.90E+03 4 6.28E-15 4 2.68E+07 4 1.52E-11 4 5.62E-12 5 0.00E+00 5 9.04E+06 5 0.00E+00 5 0.00E+00 6.5 0.00E+00 6.5 3.63E+06 6.5 0.00E+00 6.5 0.00E+00 8 0.00E+00 8 7.12E+05 8 0.00E+00 8 0.00E+00 10 0.00E+00 10 1.51E+05 10 0.00E+00 10 0.00E+00 c

c surface tallies c

c fc12 GAMMA TALLY ON TRUNNION SURFACE f12:p 868 873 fs12 -862 NT sd12 641 1E-10 641 1E-10 de12 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df12 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc14 GAMMA VOLUME TALLIES AROUND THE TRUNNION f14:p 601 602 632 662 805 sd14 4j 2.4E4 de14 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df14 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc202 GAMMA TALLY: SIDE, TOP/BOTTOM D=12', TOP/BOTTOM D=13.5', 5M f202:p 672 810 811 812 813 860 13 NT fs202 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T sd202 94215 16119 17R 15883 14983 16123 16199 72664 520209 150J de202 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df202 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc212 GAMMA TALLY: CONTACT AND 2 METERS FROM IMPACT LIMITERS f212:p 818 701 736 817 NT Page 5-52

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 fs212 -882 -883 -282 -320 -885 -886 -811 T de212 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df212 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc222 GAMMA TALLY: 1 METER SIDE ACCIDENT f222:p 814 fs222 -824 -825 -826 -827 -828 -829 -830 -831 -832

-833 -834 $ midplane segment number 11

-835 -836 -837 -838 -839 -840 -841 -842 -843 -844 -845 -846 T de222 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df222 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc232 GAMMA TALLY: 1 METER END ACCIDENT, BOTTOM, TOP f232:p 816 815 NT fs232 -882 -883 -282 -320 -885 -886 -811 T de232 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df232 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc242 GAMMA TALLY: TOP RAIL CAR EDGE AS A FUNCTION OF DISTANCE f242:p 852 853 12 $ 40', 50', 60' rail car de242 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df242 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c fc252 GAMMA TALLY: BOTTOM RAIL CAR EDGE AS A FUNCTION OF DISTANCE f252:p 851 850 11 $ 40', 50', 60' rail car de252 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 Page 5-53

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df252 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

c f24:p 906 fc24 FARTHEST TALLY FOR WWG c

fc505 GAMMA TALLY TRUNNION - on impact limiter f505:p 0. 183. 145. 2.

de505 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df505 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c

fc515 GAMMA TALLY ON TRUNNION - 2m from impact limiter f515:p 0. 383. 145. 0.25 de515 LOG 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7.5 9 11.0 13.0 15.0 df515 LOG 3.96-6 5.82-7 2.90-7 2.58-7 2.83-7 3.79-7 5.01-7 6.31-7 7.59-7 8.78-7 9.85-7 1.08-6 1.17-6 1.27-6 1.36-6 1.44-6 1.52-6 1.68-6 1.98-6 2.51-6 2.99-6 3.42-6 3.82-6 4.01-6 4.41-6 4.83-6 5.23-6 5.60-6 5.80-6 6.01-6 6.37-6 6.74-6 7.11-6 7.66-6 8.77-6 1.03-5 1.18-5 1.33-5 c control cards mode p $ gamma mode c nps 1.5E9 $ cut nps at ~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ctme 7200 prdmp j j 1 2 j $ pring end, dump every 15 min, MCTAL, keep last 2 dumps cut:p j 0.01 3j $ cut photons < 0.01 MeV bottom ANSI/ANS-6.1.1-1977 print 10 $ Source coefficients and distribution 20 $ Weight window information 30 $ Tally description

-35 $ Coincident detectors 40 $ Material composition 50 $ Cell volumes and masses, surface areas

-60 $ basic Cell importances 62 $ basic Forced collision and exponential transform

-70 $ Surface coefficients

-72 $ basic Cell temperatures

-85 $ Electron range and straggling tables multigroup: flux values for biasing adjoint calcs

-86 $ Electron bremsstrahlung and secondary production

-90 $ KCODE source data

-98 $ Physical constants and compile options 100 $ basic Cross section tables

-102 $ Assignment of S(a,b) data to nuclides

-110 $ First 50 starting histories 120 $ Analysis of the quality of your importance function Page 5-54

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 126 $ basic Particle activity in each cell

-128 $ Universe map

-130 $ Neutron/photon/electron weight balance

-140 $ Neutron/photon nuclide activity

-150 $ DXTRAN diagnostics 160 $ default TFC bin tally analysis 161 $ default f(x) tally density plot 162 $ default Cumulative f(x) and tally density plot

-170 $ Source distribution frequency tables, surface source

-175 $ shorten Estimated keff results by cycle

-178 $ Estimated keff results by batch size

-180 $ Weight window generator bookkeeping summary controlled by WWG(7), not print card

-190 $ basic Weight window generator summary

-198 $ Weight windows from multigroup fluxes

-200 $ basic Weight window generated windows c

c below is the weight window c uncomment the following to generate the weight windows c wwg 24 2072 0 4J 0 c wwge:p 8.00E-01 1.00E+00 1.50E+00 2.50E+00 $ generate WW for few key energies wwp:p 5 3 5 0 0 0 wwe:p 8.0000E-01 1.0000E+00 1.5000E+00 2.5000E+00 wwn1:p -1.0000E+00 4.7924E+02 5.3779E+02 5.5192E+02 5.0853E+02 0.0000E+00 1.3095E+02 5.0106E+03 3.1242E+02 2.4707E+02 8.3549E+01 1.8219E+01 2.0308E+01 2.0873E+01 1.9308E+01 0.0000E+00 3.9571E+01 1.8266E+02 4.6364E+01 4.5550E+01 1.9954E+01 6.3705E+00 6.3062E+00 5.8077E+00 7.9413E+00 7.1619E+00 7.9466E+00 7.2564E+00 7.6111E+00 6.7117E+00 7.8450E+00 7.0578E+00 4.6968E+00 0.0000E+00 4.9156E+00 4.2571E+00 5.1346E+00 5.3390E+00 4.2994E+00 5.0028E+00 5.2506E+00 4.4225E+00 4.8702E+00 1.4938E+00 4.5296E-01 1.3429E-01 3.8862E-02 1.1053E-02 3.0498E-03 8.7198E-04 2.4883E-04 7.1395E-05 1.1514E-02 2.5644E-03 6.7592E-04 1.7137E-04 4.5318E-05 5.1783E-06 4.5035E-06 5.6127E-06 5.1557E-06 4.1718E-05 2.1959E-05 2.1323E-05 1.8147E-05 2.9077E-05 2.7676E-05 2.6245E-05 2.4810E-05 2.3277E-05 2.1535E-05 1.9821E-05 1.8027E-05 1.6234E-05 1.4448E-05 1.2829E-05 1.1307E-05 9.8993E-06 8.7074E-06 8.0667E-06 3.8559E-06 4.3843E-06 5.8738E-06 1.1034E-05 1.2165E-05 1.2850E-05 1.2904E-05 1.2225E-05 1.0658E-05 8.8027E-06 8.1865E-06 3.5002E-06 3.6376E-06 4.3142E-06 5.5804E-06 $ go 20 lines down 6.1267E-06 6.4124E-06 6.8110E-06 7.4812E-06 2.5103E-05 $ index 100 (cell 851) 1.5414E-05 2.3034E-05 3.0355E-05 8.8634E-06 $ first cell 852 2.5103E-05 1.5414E-05 7.1215E-06 5.6559E-06 8.9884E-06 3.4899E-05 7.6275E-06 3.0673E-05 2.8070E-05 2.6039E-05 2.4335E-05 2.2723E-05 2.1213E-05 1.9606E-05 1.7988E-05 1.6350E-05 1.4718E-05 1.3132E-05 1.1621E-05 1.0226E-05 8.9371E-06 7.8589E-06 7.2944E-06 4.0325E-06 4.3934E-06 6.1579E-06 1.1024E-05 1.2454E-05 1.3056E-05 1.3197E-05 1.2509E-05 1.0693E-05 8.4180E-06 7.7788E-06 3.3673E-06 3.5672E-06 4.3165E-06 5.6553E-06 6.3306E-06 6.5654E-06 6.8204E-06 7.5660E-06 1.3241E-05 3.4861E-06 3.4627E-06 3.3930E-06 3.3872E-06 3.3793E-06 3.3755E-06 3.3494E-06 4.6824E-06 7.8162E-06 2.6969E-05 5.3477E-06 7.6486E-06 2.5721E-05 2.5658E-04 1.8583E-05 1.4697E-05 1.1209E-05 4.2679E-06 7.4331E-05 2.4992E-05 2.0928E-05 1.4992E-04 1.1982E-05 9.0481E-06 wwn2:p -1.0000E+00 2.6416E+01 3.8797E+01 3.7877E+01 3.1018E+01 0.0000E+00 1.2621E+01 3.6118E+02 2.1168E+01 9.9732E+00 Page 5-55

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.5670E+00 2.0627E+00 2.7281E+00 2.7907E+00 2.3884E+00 0.0000E+00 3.5325E+00 1.5020E+01 3.5559E+00 1.6348E+00 7.1000E-01 7.1775E-01 7.0300E-01 5.8619E-01 8.6405E-01 7.0085E-01 9.2517E-01 7.1738E-01 7.6031E-01 6.0840E-01 7.7271E-01 5.9664E-01 5.0045E-01 0.0000E+00 4.7201E-01 4.4404E-01 4.4563E-01 4.4498E-01 4.4820E-01 5.0074E-01 5.0072E-01 4.3173E-01 4.5728E-01 1.2745E-01 3.8197E-02 1.1817E-02 3.8262E-03 1.2746E-03 4.0751E-04 1.4571E-04 5.4138E-05 2.1296E-05 1.6465E-03 4.0549E-04 1.3140E-04 4.5056E-05 1.7332E-05 4.1175E-06 3.9497E-06 4.4253E-06 4.2298E-06 1.7141E-05 1.0075E-05 1.1089E-05 9.9781E-06 1.1887E-05 1.0936E-05 1.0015E-05 9.2021E-06 8.4453E-06 7.7606E-06 7.1606E-06 6.6202E-06 6.1256E-06 5.7118E-06 5.3139E-06 4.9852E-06 4.7158E-06 4.5076E-06 4.4078E-06 3.8325E-06 3.9727E-06 4.3122E-06 4.8212E-06 4.8904E-06 4.8052E-06 4.5252E-06 4.1552E-06 3.7382E-06 3.3452E-06 3.2759E-06 3.7920E-06 3.7454E-06 3.7475E-06 3.5633E-06 3.5230E-06 3.4854E-06 3.1516E-06 3.1170E-06 1.0246E-05 6.6142E-06 9.7516E-06 1.1673E-05 4.5474E-06 1.0246E-05 6.6142E-06 4.7554E-06 4.3276E-06 5.1422E-06 1.4238E-05 4.7855E-06 1.2774E-05 1.1622E-05 1.0557E-05 9.5956E-06 8.7578E-06 7.9689E-06 7.2828E-06 6.6793E-06 6.1455E-06 5.6668E-06 5.2609E-06 4.8909E-06 4.5505E-06 4.2734E-06 4.0542E-06 3.9728E-06 3.8031E-06 3.9959E-06 4.3156E-06 4.4527E-06 4.6597E-06 4.7678E-06 4.4195E-06 4.1549E-06 3.6316E-06 3.0454E-06 2.9702E-06 3.5527E-06 3.6378E-06 3.7397E-06 3.5493E-06 3.6009E-06 3.5652E-06 3.2133E-06 3.1647E-06 7.0835E-06 3.6583E-06 3.6610E-06 3.6630E-06 3.6694E-06 3.6775E-06 3.6769E-06 3.6611E-06 5.8741E-06 1.0121E-05 3.8117E-05 6.7098E-06 9.6910E-06 3.5156E-05 7.0642E-05 1.0956E-05 8.9196E-06 1.6047E-05 4.1713E-06 4.4477E-05 2.5318E-05 6.8572E-06 3.9162E-06 5.6294E-06 4.4523E-06 wwn3:p -1.0000E+00 3.7412E+00 4.4051E+00 4.5375E+00 4.1216E+00 0.0000E+00 1.9031E+00 3.8071E+01 3.9090E+00 1.5734E+00 5.7994E-01 3.7082E-01 4.4512E-01 4.5354E-01 4.0996E-01 0.0000E+00 7.1811E-01 2.1849E+00 7.1860E-01 4.1150E-01 2.0096E-01 1.2953E-01 1.2678E-01 1.1017E-01 1.4740E-01 1.2322E-01 1.4709E-01 1.2445E-01 1.5023E-01 1.2623E-01 1.4799E-01 1.2372E-01 8.9336E-02 0.0000E+00 8.9329E-02 8.1925E-02 9.0412E-02 8.8926E-02 8.5167E-02 8.7213E-02 8.8982E-02 8.2431E-02 1.0734E-01 3.1554E-02 1.0512E-02 3.7185E-03 1.3736E-03 5.2603E-04 1.9555E-04 8.1317E-05 3.5173E-05 1.6101E-05 6.8938E-04 1.9817E-04 7.6316E-05 3.1109E-05 1.3693E-05 3.9609E-06 3.9028E-06 4.2673E-06 4.1051E-06 1.3515E-05 8.5848E-06 9.4557E-06 8.6183E-06 9.9399E-06 9.3299E-06 8.7134E-06 8.1674E-06 7.6481E-06 7.1464E-06 6.7080E-06 6.3102E-06 5.9766E-06 5.6563E-06 5.4267E-06 5.2062E-06 5.0347E-06 4.9479E-06 4.9588E-06 3.8518E-06 3.9072E-06 4.1569E-06 4.3849E-06 4.3098E-06 4.2963E-06 4.0815E-06 3.7999E-06 3.4336E-06 3.1335E-06 3.0967E-06 3.8454E-06 3.7717E-06 3.7471E-06 3.4974E-06 3.3627E-06 3.2514E-06 2.9471E-06 2.9861E-06 8.4317E-06 5.9163E-06 9.2380E-06 1.0535E-05 4.4103E-06 8.4317E-06 5.9163E-06 4.4073E-06 4.1476E-06 4.6954E-06 1.1214E-05 4.4535E-06 1.0293E-05 9.5178E-06 8.7972E-06 8.1605E-06 7.5961E-06 7.0904E-06 6.6341E-06 6.1913E-06 5.8101E-06 5.4882E-06 5.1925E-06 4.9219E-06 4.7084E-06 4.5250E-06 4.3982E-06 4.3778E-06 3.8378E-06 3.9422E-06 4.1544E-06 4.1276E-06 4.3770E-06 4.4493E-06 4.2162E-06 3.8786E-06 3.3747E-06 2.8683E-06 Page 5-56

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 2.8162E-06 3.6744E-06 3.7568E-06 3.7754E-06 3.5147E-06 3.3655E-06 3.3647E-06 2.9433E-06 2.9010E-06 5.4682E-06 3.7632E-06 3.7647E-06 3.7666E-06 3.7680E-06 3.8004E-06 3.8010E-06 3.7612E-06 6.5034E-06 1.0981E-05 4.0803E-05 7.8400E-06 1.1113E-05 4.3453E-05 7.0180E-05 1.0642E-05 7.8058E-06 6.5446E-06 4.1026E-06 3.1959E-05 1.4227E-05 8.9483E-06 0.0000E+00 4.7515E-06 4.3386E-06 wwn4:p -1.0000E+00 1.1128E+00 1.2719E+00 1.3126E+00 1.2030E+00 0.0000E+00 4.6928E-01 7.6998E+00 7.1310E-01 3.4324E-01 1.2987E-01 1.0612E-01 1.2452E-01 1.2690E-01 1.1606E-01 0.0000E+00 1.8632E-01 5.0000E-01 1.6127E-01 9.7304E-02 5.0868E-02 3.7377E-02 3.6663E-02 3.2286E-02 4.3562E-02 3.7005E-02 4.3381E-02 3.7706E-02 4.2230E-02 3.6518E-02 4.2808E-02 3.7119E-02 2.6831E-02 0.0000E+00 2.6548E-02 2.3994E-02 2.6521E-02 2.7063E-02 2.5741E-02 2.6688E-02 2.6943E-02 2.3856E-02 3.0825E-02 9.9530E-03 3.6923E-03 1.4656E-03 6.0753E-04 2.6090E-04 1.1172E-04 5.1763E-05 2.4938E-05 1.2726E-05 3.0081E-04 1.0147E-04 4.6041E-05 2.2013E-05 1.1000E-05 3.8639E-06 3.8445E-06 4.0534E-06 3.9424E-06 9.8329E-06 7.2038E-06 8.1545E-06 7.6196E-06 8.5680E-06 8.1971E-06 7.8889E-06 7.5370E-06 7.1450E-06 6.7558E-06 6.4702E-06 6.2040E-06 5.9969E-06 5.7971E-06 5.6377E-06 5.5396E-06 5.4782E-06 5.5281E-06 5.5828E-06 3.8331E-06 3.8206E-06 3.9923E-06 4.1548E-06 4.0861E-06 3.7829E-06 3.5133E-06 3.3025E-06 3.1857E-06 3.0618E-06 3.1082E-06 3.8541E-06 3.7349E-06 3.6721E-06 3.4335E-06 3.3773E-06 3.1450E-06 2.9953E-06 3.0143E-06 7.1726E-06 5.3780E-06 7.8604E-06 9.9838E-06 4.2522E-06 7.1726E-06 5.3780E-06 4.1443E-06 4.0177E-06 4.3654E-06 9.2171E-06 4.2447E-06 8.6710E-06 8.0942E-06 7.5382E-06 7.1205E-06 6.7122E-06 6.3400E-06 6.0123E-06 5.7331E-06 5.4615E-06 5.2318E-06 5.0211E-06 4.8093E-06 4.6441E-06 4.5070E-06 4.4702E-06 4.5198E-06 3.8563E-06 3.9131E-06 4.0408E-06 3.7396E-06 3.7333E-06 3.7239E-06 3.9499E-06 3.6815E-06 3.2540E-06 2.7409E-06 2.7449E-06 3.7520E-06 3.7808E-06 3.7851E-06 3.4067E-06 3.2798E-06 3.1959E-06 2.7225E-06 2.7188E-06 4.6654E-06 3.8159E-06 3.8167E-06 3.8182E-06 3.8137E-06 3.8586E-06 3.8595E-06 3.8123E-06 7.0550E-06 1.1635E-05 4.7361E-05 9.5474E-06 1.4248E-05 5.3783E-05 1.1840E-04 1.0324E-05 1.8482E-05 0.0000E+00 3.9363E-06 2.7006E-05 4.7021E-06 1.0683E-05 0.0000E+00 4.4044E-06 4.0063E-06 Page 5-57

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-1 TN-40 Cask Shield Materials Density Thickness Component Material (g/cm3) (inches)

Cask Body Wall Carbon Steel 7.82 9.50 Lid Carbon Steel 7.82 10.50 Bottom Carbon Steel 7.82 10.25 Polyester Resin Styrene Resin a 1.58 4.50 Aluminum Hydrate Zinc Borate Aluminum Box Aluminum 2.7 0.12 Outer Shell Carbon Steel 7.82 0.50 Stainless Steel (fuel 7.94 0.09 compartment)

Basket Aluminum 2.7 0.25 x 2 Neutron Poison Materialb 2.45 0.075 Stainless Steel 7.94 0.25 Impact Limiter Redwood 0.387 35.0c Balsa Wood 0.125 2.5c Notes:

a ` The neutron shielding is borated polyester resin compound with a density of 1.58 g/cc. The four major constituents are listed in the table.

b ` This is modeled as aluminum for shielding purposes with a reduced density.

c ` Thickness of wood is variable.

Page 5-58

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-2 Summary of TN-40 Dose Rates (Exclusive Use)

Normal Package Contact Dose Rate 2 Meters from Vehicle Surface Conditions of mSv/hour (mrem/hour) mSv/hour (mrem/hour)

Transport Vehicle Surface(1)

Radiation Top Side Bottom mSv/hour (mrem/hour) Top Side Bottom 0.069 0.30 0.059 0.19 0.035 Gamma - -

(6.9) (30) (5.9) (19) (3.5) 0.0004 0.30 0.0008 0.21 0.043 Neutron - -

(0.04) (30) (0.08) (21) (4.3) 0.069 0.60 0.060 0.40 <0.069 0.078 <0.060 Total (6.9) (60) (6.0) (40) (<6.9) (7.8) (<6.0) 2 10 2 2 Limit 0.1 (10) 0.1 (10) 0.1 (10)

(200) (1000) (200) (200)

(1) Vehicle surface is bounded axially by the external surfaces of the impact limiters and radially by the vertical planes extending from a 10 ft wide vehicle. The bounding radial dose rates are shown for all surfaces.

Hypothetical 1 Meter from Package Surface Accident Conditions(2) mSv/hour (mrem/hour)

Radiation Top Side(3) Bottom Gamma 0.43 (43) 0.32 (32) 0.28 (28)

Neutron 0.68 (68) 5.34 (534) 1.45 (145)

Total 1.11 (111) 5.66 (566) 1.73 (173)

Limit 10 (1000) 10 (1000) 10 (1000)

(2) The neutron shield and the impact limiters are removed.

(3) Does not account for tolerances on side of the cask described at the end of Section 5.1. The effect of tolerances is less than 10%. It is not significant to the extent that dose rates would exceed regulatory limits.

Page 5-59

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-3 PWR Fuel Assembly Design Characteristics Westinghouse Parameter Standard (14x14)

Max Length (in) 161.1 Max Width (in) 7.763 Rod Pitch (in) 0.556 No of Fueled Rods 179 Fuel Rod Length (in) 152 Maximum Active Fuel Length (in) 144.0 Fuel Rod OD (in) 0.4220 Clad Thickness (in) 0.0243 Fuel Pellet OD (in) 0.3659 Clad Material Zr-4 Guide Tube OD (in) 0.539 Guide Tube Wall Thickness (in) 0.017 Guide Tube # 16 Instrument Tube # 1 Instr. Tube OD (in) 0.422 Instr. Tube Wall Thickness (in) 0.0243 Maximum MTU/assembly 0.410 Page 5-60

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-4 Westinghouse 14 X 14 STD Fuel Assembly Hardware Characteristics Average Mass Item Material (kg/assembly)

Fuel Zone Cladding Zircaloy 83.4 Spacers Inconel 5.37 Guide & Instrument Tubes Stainless Steel(1) 7.74 Fuel-Gas Plenum Zone Cladding Zircaloy 4.13 Springs Stainless Steel 5.68 Guide & Instrument Tubes Stainless Steel(1) 0.38 Spacer Inconel 0.68 Top End Fitting Zone Top Nozzle Stainless Steel 6.30 Hold Down Springs Inconel 0.51 Bottom End Fitting Zone Bottom Nozzle Stainless Steel 7.89 Total 122.0 (1) The zircaloy guide and instrument tubes are modeled as stainless steel to include the effect of BPRAs in the source term calculations.

Page 5-61

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-5 Material Compositions for Fuel Assembly Hardware Materials Material Composition, grams per kg of material Atomic Element Inconel X- Stainless U02 Fuel Number Zircaloy-4 Inconel-718 750 Steel 304 (per kg U)

H 1 1.30E-02 - - - -

Li 3 - - - - 1.00E-03 B 5 3.30E-04 - - - 1.00E-03 C 6 1.20E-01 4.00E-01 3.99E-01 8.00E-01 8.94E-02 N 7 8.00E-02 1.30E+00 1.30E+00 1.30E+00 2.50E-02 O 8 9.50E-01 - - - 1.34E+02 F 9 - - - - 1.07E-02 Na 11 - - - - 1.50E-02 Mg 12 - - - - 2.00E-03 Al 13 2.40E-02 5.99E+00 7.98E+00 - 1.67E-02 Si 14 - 2.00E+00 2.99E+00 1.00E+01 1.21E-02 P 15 - - - 4.50E-01 3.50E-02 S 16 3.50E-02 7.00E-02 7.00E-02 3.00E-01 -

Cl 17 - - - - 5.30E-03 Ca 20 - - - - 2.00E-03 Ti 22 2.00E-02 7.99E+00 2.49E+01 - 1.00E-03 V 23 2.00E-02 - - - 3.00E-03 Cr 24 1.25E+00 1.90E+02 1.50E+02 1.90E+02 4.00E-03 Mn 25 2.00E-02 2.00E+00 6.98E+00 2.00E+01 1.70E-03 Fe 26 2.25E+00 1.80E+02 6.78E+01 6.88E+02 1.80E-02 Co 27 1.00E-02 4.69E+00 6.49E+00 8.00E-01 1.00E-03 Ni 28 2.00E-02 5.20E+02 7.22E+02 8.92E+01 2.40E-02 Cu 29 2.00E-02 9.99E-01 4.99E-01 - 1.00E-03 Zn 30 - - - - 4.03E-02 Zr 40 9.79E+02 - - - -

Nb 41 - 5.55E+01 8.98E+00 - -

Mo 42 - 3.00E+01 - - 1.00E-02 Ag 47 - - - - 1.00E-04 Cd 48 2.50E-04 - - - 2.50E-02 In 49 - - - - 2.00E-03 Sn 50 1.60E+01 - - - 4.00E-03 Gd 64 - - - - 2.50E-03 Hf 72 7.80E-02 - - - -

W 74 2.00E-02 - - - 2.00E-03 Pb 82 - - - - 1.00E-03 U 92 2.00E-04 - - - 1.00E+03 Page 5-62

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-6 Source Distribution Source[1] (particles/sec/assembly)

Elower Eupper Bottom Fitting Active Fuel Plenum Top Fitting Combined (MeV) (MeV) Gamma Gamma Gamma Gamma Gamma 0.01 0.05 1.15E+10 4.82E+14 1.58E+10 7.03E+09 4.82E+14 0.05 0.10 2.19E+09 1.41E+14 2.81E+09 1.29E+09 1.41E+14 0.10 0.20 5.27E+08 8.83E+13 7.15E+08 3.11E+08 8.83E+13 0.20 0.30 2.62E+07 2.70E+13 3.61E+07 1.55E+07 2.70E+13 0.30 0.40 3.43E+07 1.81E+13 5.27E+07 2.02E+07 1.81E+13 0.40 0.60 2.17E+06 1.47E+13 1.94E+08 1.28E+06 1.47E+13 0.60 0.80 7.54E+05 9.70E+14 8.56E+08 2.83E+08 9.70E+14 0.80 1.00 2.84E+07 8.23E+12 7.65E+08 2.89E+08 8.23E+12 1.00 1.33 6.38E+11 2.75E+13 2.17E+12[2] 1.29E+12[2] 3.16E+13[2]

1.33 1.66 1.80E+11 5.38E+12 6.11E+11[2] 3.63E+11[2] 6.54E+12[2]

1.66 2.00 1.64E-06 4.63E+10 1.47E+01 1.10E-03 4.63E+10 2.00 2.50 4.27E+06 2.48E+09 5.48E+06 2.51E+06 2.50E+09 2.50 3.00 6.63E+03 1.73E+08 8.50E+03 3.90E+03 1.73E+08 3.00 4.00 6.28E-15 2.68E+07 1.52E-11 5.62E-12 2.68E+07 4.00 5.00 0.00E+00 9.04E+06 0.00E+00 0.00E+00 9.04E+06 5.00 6.50 0.00E+00 3.63E+06 0.00E+00 0.00E+00 3.63E+06 6.50 8.00 0.00E+00 7.12E+05 0.00E+00 0.00E+00 7.12E+05 8.00 10.00 0.00E+00 1.51E+05 0.00E+00 0.00E+00 1.51E+05 Total Gamma 8.32E+11 1.78E+15 2.80E+12 1.66E+12 1.79E+15 Total Neutron[3] 2.63E+08

1. The design basis gamma source term correspond to the Westinghouse 14x14 Standard fuel assembly with 2.35 wt.% U-235 enrichment , 42,000 MWD/MTU burnup, 24.4 year cooling time and TPA insert source term.

2. Total gamma source from the fuel assembly and the TPA source shown in Table 5-7.

3. The neutron source spectrum is modeled as Cm-244 using the built-in MCNP distribution.

Table 5-7 TPA Gamma Source Gamma Source from TPA (gammas/sec per assembly)

Plenum Top End Fitting Total 1 - 1.33 MeV 1.33 - 1.66 MeV 1 - 1.33 MeV 1.33 - 1.66 MeV 1.35E+12 3.81E+11 9.13E+11 2.57E+11 2.90E+12 Page 5-63

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-8 Fuel Qualification Table Minimum Cooling Times (Years)

Maximum Assembly Minimum Assembly Average Enrichment Average Burnup (wt. % U235)

(GWD/MTU) 2 2.25 2.35 2.75 3 3.25 3.4 3.6 3.85 17 15 15 15 15 15 15 15 15 15 18 15 15 15 15 15 15 15 15 15 19 15 15 15 15 15 15 15 15 15 20 15 15 15 15 15 15 15 15 15 21 15 15 15 15 15 15 15 15 15 22 15 15 15 15 15 15 15 15 15 23 15 15 15 15 15 15 15 15 15 24 15 15 15 15 15 15 15 15 15 25 15 15 15 15 15 15 15 15 15 26 15 15 15 15 15 15 15 15 15 27 15 15 15 15 15 15 15 15 15 28 15 15 15 15 15 15 15 15 15 29 15 15 15 15 15 15 15 30 15 15 15 15 15 15 15 31 15 15 15 15 15 15 15 32 15 15 15 15 15 15 15 33 16 15 15 15 15 15 15 34 17 15 15 15 15 15 15 35 17 16 15 15 15 15 15 36 18 16 15 15 15 15 15 37 19 17 16 15 15 15 15 38 20 18 17 16 16 15 15 39 21 19 18 17 16 16 15 40 23 20 19 18 17 16 16 41 24 21 20 19 18 17 17 42 25 22 21 19 19 18 18 43 22 20 20 20 19 44 21 21 21 21 45 23 22 22 22 Notes:

1. For fuel characteristics that fall between the minimum assembly average enrichment values in the table, use the next lower enrichment, and next higher burnup to determine minimal cooling time.

2. Enrichment and burnup are also required to meet criticality requirements as defined in Figure 6-1. The minimum required cooling time to meet the criticality analysis requirements is 30 years.

Page 5-64

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-9 Minimum Cooling (Years) Required to Meet Radiation and Decay Heat Limits Maximum Assembly Minimum Assembly Average Initial Enrichment Average Burnup (Wt. % U235)

(GWD/MTU) 2 2.25 2.35 2.75 3 3.25 3.4 3.6 3.85 17 7.4 7.0 18 7.8 7.4 19 8.3 7.8 20 8.7 8.2 21 9.2 8.7 22 9.7 9.2 23 10.3 9.7 24 10.9 10.2 25 10.9 10.2 26 11.5 10.8 10.5 27 12.1 11.4 11.1 9.6 9.2 28 12.8 12.0 11.7 10.6 10.1 9.6 29 12.3 11.2 10.6 10.1 9.8 30 13.0 11.7 11.1 10.6 10.3 31 13.6 12.3 11.7 11.1 10.8 32 14.4 13.0 12.3 11.6 11.3 33 15.2 13.7 12.9 12.2 11.8 11.4 34 16.0 14.4 13.5 12.8 12.4 11.9 11.5 35 16.8 15.1 14.2 13.4 13.0 12.5 12.0 36 17.8 15.9 14.9 14.1 13.6 13.1 12.6 37 18.7 16.7 15.7 14.8 14.3 13.7 13.2 38 19.8 17.6 16.5 15.5 15.0 14.4 13.8 39 20.9 18.5 17.3 16.3 15.7 15.0 14.4 40 22.0 19.5 18.2 17.1 16.5 15.7 15.1 41 23.2 20.5 19.2 18.0 17.3 16.5 16.3 42 24.4 21.7 20.2 18.9 18.2 17.8 17.5 43 21.2 19.8 19.2 19.0 18.9 44 20.8 20.6 20.4 20.2 45 22.1 21.9 21.7 21.6 Page 5-65

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-10 Dose Rates (mrem/hr) at 2 Meters from Side of 10 wide Transportation Platform Estimated with Response Function Maximum Assembly Minimum Assembly Average Initial Enrichment Average Burnup (Wt. % U235)

(GWD/MTU) 2 2.25 2.35 2.75 3 3.25 3.4 3.6 3.85 17 10 10 18 10 10 19 10 10 20 10 10 21 10 10 22 10 10 23 10 10 24 10 10 25 10 10 26 10 10 10 27 10 10 10 10 10 28 10 10 10 10 10 10 29 10 10 10 10 10 30 10 10 10 10 10 31 10 10 10 10 10 32 10 10 10 10 10 33 10 10 10 10 10 10 34 10 10 10 10 10 10 10 35 10 10 10 10 10 10 10 36 10 10 10 10 10 10 10 37 10 10 10 10 10 10 10 38 10 10 10 10 10 10 10 39 10 10 10 10 10 10 10 40 10 10 10 10 10 10 10 41 10 10 10 10 10 10 9.4 42 10 10 10 10 10 9.4 9.1 43 10 10 9.6 9.2 8.7 44 10 9.4 8.9 8.5 45 9.6 9.2 8.7 8.2 Page 5-66

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-11 Decay Heat Output (kW PER CASK)

Maximum Assembly Minimum Assembly Average Initial Enrichment A\verage Burnup (Wt. % U235)

(GWD/MTU) 2 2.25 2.35 2.75 3 3.25 3.4 3.6 3.85 17 11.4 11.8 18 11.8 12.1 19 12.1 12.4 20 12.4 12.8 21 12.8 13.1 22 13.1 13.4 23 13.4 13.7 24 13.8 14.1 25 14.1 14.4 26 14.4 14.7 14.9 27 14.8 15.1 15.2 16.0 16.2 28 15.1 15.4 15.5 16.0 16.3 16.6 29 15.8 16.3 16.6 16.9 17.1 30 16.1 16.7 16.9 17.2 17.4 31 16.5 17.0 17.3 17.6 17.7 32 16.7 17.3 17.6 17.9 18.0 33 17.0 17.6 17.9 18.2 18.4 18.6 34 17.3 17.8 18.2 18.5 18.7 18.9 19.1 35 17.6 18.2 18.5 18.8 19.0 19.2 19.5 36 17.8 18.5 18.8 19.1 19.3 19.6 19.8 37 18.0 18.7 19.1 19.4 19.6 19.8 20.1 38 18.2 18.9 19.3 19.7 19.9 20.1 20.4 39 18.4 19.2 19.6 19.9 20.2 20.4 20.7 40 18.5 19.4 19.8 20.2 20.4 20.7 20.9 41 18.6 19.5 20.0 20.4 20.6 21.0 21.0 42 18.7 19.6 20.1 20.6 20.8 20.9 21.0 43 20.3 20.8 21.0 21.0 21.0 44 21.0 21.0 21.0 21.0 45 21.0 21.0 21.0 21.0 Page 5-67

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-12 Axial Source Term Peaking Summary Gamma Neutron Fractional Core Height Profile Profile 0 to 0.083 0.700 0.240 0.083 to 0.167 1.030 1.126 0.167 to 0.250 1.120 1.574 0.250 to 0.333 1.160 1.811 0.333 to 0.417 1.160 1.811 0.417 to 0.500 1.160 1.811 0.500 to 0.583 1.158 1.798 0.583 to 0.667 1.150 1.749 0.667 to 0.750 1.144 1.713 0.750 to 0.833 1.120 1.574 0.833 to 0.917 0.990 0.961 0.917 to 1.000 0.700 0.240 Average: 1.049 1.367 Page 5-68

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-13 Fuel Assembly Materials Input For MCNP (2 Sheets)

Lower End Fitting Zone Mass Density Weight Number Density Atom Element/Isotope (g/cc) Fraction (atom/barn-cm) Fraction C 2.08E-03 0.080% 1.04E-04 0.364%

Si 2.59E-02 1.000% 5.56E-04 1.944%

P 1.17E-03 0.045% 2.27E-05 0.079%

Cr 4.93E-01 19.000% 5.71E-03 19.949%

Ti 0.00E+00 0.000% 0.00E+00 0.000%

Mn 5.19E-02 2.000% 5.69E-04 1.987%

Fe 1.77E+00 68.375% 1.91E-02 66.841%

Ni 2.47E-01 9.500% 2.53E-03 8.836%

Zr 0.00E+00 0.000% 0.00E+00 0.000%

Sn 0.00E+00 0.000% 0.00E+00 0.000%

Hf 0.00E+00 0.000% 0.00E+00 0.000%

O 0.00E+00 0.000% 0.00E+00 0.000%

H 0.00E+00 0.000% 0.00E+00 0.000%

TOTAL 2.595 100.0% 0.02863 100.0%

In-Core Zone Mass Density Weight Number Density Atom Element/Isotope (g/cc) Fraction (atom/barn-cm) Fraction U-234 7.70E-04 0.019% 1.98E-06 0.007%

U-235 8.65E-02 2.190% 2.22E-04 0.828%

U-236 3.98E-04 0.010% 1.02E-06 0.004%

U-238 2.80E+00 70.788% 7.07E-03 26.424%

O 3.88E-01 9.815% 1.46E-02 54.526%

C 4.35E-05 0.0011% 2.18E-06 0.008%

Si 1.49E-03 0.038% 3.19E-05 0.119%

P 2.45E-05 0.0006% 4.76E-07 0.0018%

Cr 1.66E-02 0.420% 1.92E-04 0.718%

Ti 9.44E-04 0.024% 1.19E-05 0.044%

Mn 1.09E-03 0.028% 1.19E-05 0.045%

Fe 4.11E-02 1.040% 4.43E-04 1.655%

Ni 3.27E-02 0.829% 3.36E-04 1.255%

Zr 5.76E-01 14.581% 3.80E-03 14.203%

Sn 8.50E-03 0.215% 4.31E-05 0.161%

Hf 5.86E-05 0.001% 1.98E-07 0.001%

H 0.00E+00 0.000% 0.00E+00 0.000%

TOTAL 3.949 100.0% 0.02676 100.0%

Page 5-69

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-13 Fuel Assembly Materials Input For MCNP (2 Sheets)

Plenum Zone Mass Density Weight Number Density Atom Element/Isotope (g/cc) Fraction (atom/ barn-cm) Fraction C 6.88E-04 0.045% 3.45E-05 0.240%

Si 1.10E-02 0.714% 2.36E-04 1.640%

P 3.87E-04 0.025% 7.53E-06 0.052%

Cr 1.78E-01 11.569% 2.07E-03 14.358%

Ti 2.41E-03 0.156% 3.03E-05 0.211%

Mn 1.72E-02 1.115% 1.89E-04 1.310%

Fe 5.96E-01 38.641% 6.43E-03 44.650%

Ni 1.52E-01 9.859% 1.56E-03 10.839%

Zr 5.76E-01 37.321% 3.80E-03 26.400%

Sn 8.50E-03 0.551% 4.31E-05 0.299%

Hf 5.86E-05 0.004% 1.98E-07 0.001%

O 0.00E+00 0.000% 0.00E+00 0.000%

H 0.00E+00 0.000% 0.00E+00 0.000%

TOTAL 1.543 100.0% 0.01440 100.0%

Upper End Fitting Zone Mass Density Weight Number Density Atom Element/Isotope (g/cc) Fraction (atom/barn-cm) Fraction C 1.46E-03 0.074% 7.31E-05 0.337%

Si 2.19E-02 1.112% 4.70E-04 2.164%

P 8.20E-04 0.042% 1.59E-05 0.073%

Cr 3.68E-01 18.702% 4.27E-03 19.662%

Ti 3.67E-03 0.187% 4.62E-05 0.213%

Mn 3.65E-02 1.851% 4.00E-04 1.842%

Fe 1.26E+00 63.795% 1.36E-02 62.448%

Ni 2.80E-01 14.238% 2.88E-03 13.261%

Zr 0.00E+00 0.000% 0.00E+00 0.000%

Sn 0.00E+00 0.000% 0.00E+00 0.000%

Hf 0.00E+00 0.000% 0.00E+00 0.000%

O 0.00E+00 0.000% 0.00E+00 0.000%

H 0.00E+00 0.000% 0.00E+00 0.000%

TOTAL 1.970 100.0% 0.02170 100.0%

Page 5-70

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-14 Package Materials Input for MCNP Weight Density Element/ Library Fraction Zone Material (g/cc) Nuclide Identifier (atm fraction)

Cr 24000 0.1900 Mn 25000 0.0200 Basket Plates Fe 26000 0.68375

& Impact SS304 7.94 Ni 28000 0.0950 Limiter Skin Si 14000 0.0100 P 15031 0.00045 C 6012 0.00080 Basket Plates Aluminum 2.702 Al 13027 1.0000

& Rails Fe 26000 0.9900 Cask Body Carbon Steel 7.8212 C 6012 0.0100 O 8016 0.3503 Al 13027 0.2851 Resin (1.58 g/cc) C 6012 0.2953 Resin/Alumin

& 1.687 H 1001 0.04260 um Al (2.702 g/cc) B-10 5010 0.0018 B-11 5011 0.0071 Zn 30000 0.0178 C 6012 (0.2857)

Impact Balsa Wood 0.125 O 8016 (0.2381)

Limiter H 1001 (0.4762)

C 6012 (0.2857)

Impact Redwood 0.387 O 8016 (0.2381)

Limiter H 1001 (0.4762)

Page 5-71

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-15 Flux-to-Dose Rate Conversion Factors for Gamma Photon Conversion Energy Factor (MeV) (rem/hr) / (/cm2-s) 0.01 3.96E-06 0.03 5.82E-07 0.05 2.90E-07 0.07 2.58E-07 0.1 2.83E-07 0.15 3.79E-07 0.2 5.01E-07 0.25 6.31E-07 0.3 7.59E-07 0.35 8.78E-07 0.4 9.85E-07 0.45 1.08E-06 0.5 1.17E-06 0.55 1.27E-06 0.6 1.36E-06 0.65 1.44E-06 0.7 1.52E-06 0.8 1.68E-06 1 1.98E-06 1.4 2.51E-06 1.8 2.99E-06 2.2 3.42E-06 2.6 3.82E-06 2.8 4.01E-06 3.25 4.41E-06 3.75 4.83E-06 4.25 5.23E-06 4.75 5.60E-06 5 5.80E-06 5.25 6.01E-06 5.75 6.37E-06 6.25 6.74E-06 6.75 7.11E-06 7.5 7.66E-06 9 8.77E-06 11 1.03E-05 13 1.18E-05 15 1.33E-05 Page 5-72

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-16 Flux-Dose-Rate Conversion Factors for Neutron Neutron Conversion Energy Factor (MeV) (rem/hr) / (n/cm2-s) 2.50E-08 3.67E-06 1.0E -07 3.67E-06 1.00E-06 4.46E-06 1.00E-05 4.54E-06 1.00E-04 4.18E-06 1.00E-03 3.76E-06 1.00E-02 3.56E-06 1.00E-01 2.17E-05 5.00E-01 9.26E-05 1 1.32E-04 2.5 1.25E-04 5 1.56E-04 7 1.47E-04 10 1.47E-04 14 2.08E-04 20 2.27E-04 Page 5-73

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-17 Average End Dose Rates as a Function of Railcar Length Rail Car Average Dose Rates as a Function of Railcar Length (mrem/hr)

Length Neutron N, Gamma Gamma Total (feet) Limit Top Bottom Top Bottom Top Bottom Top Bottom 40 0.0394 0.0410 0.0453 0.0769 0.686 0.425 0.770 0.543 50 0.0329 0.0325 0.0246 0.0433 0.408 0.252 0.466 0.328 2 60 0.0607 0.0540 0.0202 0.0266 0.107 0.078 0.188 0.159 Page 5-74

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-18 Total Dose Rates along Side of the Cask at Normal Conditions-Design Basis Axial Range (cm). Note, (1)

Total Dose Rates ends of impact limiters (mrem/hr) are at 2 Meter from approximately -312 cm Vertical Bottom/ 2 Meter from edge of impact and 346 cm (2)

Contact Side Top 120 vehicle (3) limiter (4)

-306.9 to -190 19.7 1.36 1.90 1.88 1.21

-190 to -170 48.0 28.4 16.7 6.24 5.60

-170 to -150 26.2 23.5 19.0 7.05 6.27

-150 to -130 31.0 24.4 20.1 7.71 6.87

-130 to -110 36.8 27.4 21.6 8.31 7.35

-110 to -90 41.3 30.2 23.4 8.83 7.85

-90 to -70 44.0 32.2 24.7 9.36 8.17

-70 to -50 45.5 33.2 25.9 9.65 8.45

-50 to -30 46.5 34.2 26.4 9.96 8.67

-30 to -10 47.3 34.0 26.8 10.10 8.87

-10 to 10 46.6 34.5 26.6 10.20 9.04 10 to 30 46.5 33.9 26.6 10.30 9.05 30 to 50 46.1 33.8 26.2 10.30 9.03 50 to 70 45.5 33.1 25.8 10.20 9.06 70 to 90 42.9 31.3 24.5 10.30 8.99 90 to 110 39.9 29.3 23.2 10.10 8.89 110 to 130 35.3 27.0 22.2 9.97 8.68 130 to 150 31.6 24.9 21.4 9.66 8.50 150 to 170 27.7 24.8 22.3 9.44 8.36 170 to 190 30.4 29.2 25.6 9.25 8.07 190 to 210 72.6 49.1 32.0 8.90 7.82 210 to 230 137 61.1 30.3 8.30 7.34 230 to 250 41.4 20.6 9.65 7.36 6.54 250 to 340.16 6.55 1.11 1.79 1.94 1.28 Maximum Relative Error 1.9% 1.2% 1.8% 1.1% 1.1%

(1) Accounts for cumulative effect of tolerances (+.05/-.01 on 1.50 thick inner shell and +/-.12 on 8.00 thick gamma shell) of steel thicknesses and tolerances (+/-.12 on 4.50) of resin on side of the cask.

(2) Contact dose rates at axial coordinates less than -187 cm or greater than 220 cm (in bolded cells of the table) are inside of impact limiters.

(3) The maximum 2 m dose rate at the axis of the trunnion (z=170 cm) is 9.50 mrem/hour.

(4) The maximum 2 m dose rate at the axis of the trunnion (z=170 cm) is 8.76 mrem/hour.

Page 5-75

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-19 Total Dose Rates along Side of the Cask at Normal Conditions-30 Year Cooled Axial Range (cm). Note, (1)

Total Dose Rates ends of impact limiters (mrem/hr) are at 2 Meter from approximately -312 cm Vertical Bottom/ 2 Meter from edge of impact and 346 cm (2)

Contact Side Top 120 vehicle (3) limiter (4)

-306.9 to -190 15.1 0.92 1.33 1.42 0.91

-190 to -170 37.0 21.6 12.5 4.66 4.19

-170 to -150 19.4 17.7 14.3 5.30 4.69

-150 to -130 23.3 18.3 15.1 5.79 5.14

-130 to -110 27.5 20.5 16.2 6.25 5.51

-110 to -90 31.1 22.7 17.7 6.63 5.88

-90 to -70 33.4 24.3 18.6 7.03 6.13

-70 to -50 34.6 25.2 19.6 7.24 6.33

-50 to -30 35.5 25.9 20.0 7.50 6.51

-30 to -10 36.1 25.8 20.3 7.59 6.67

-10 to 10 35.4 26.2 20.1 7.70 6.80 10 to 30 35.4 25.7 20.1 7.78 6.82 30 to 50 35.0 25.6 19.9 7.79 6.83 50 to 70 34.6 25.0 19.4 7.71 6.85 70 to 90 32.3 23.5 18.4 7.79 6.82 90 to 110 30.0 21.9 17.4 7.71 6.78 110 to 130 26.5 20.2 16.7 7.62 6.66 130 to 150 23.4 18.7 16.3 7.44 6.56 150 to 170 20.9 19.0 17.3 7.33 6.46 170 to 190 23.9 23.1 20.3 7.20 6.29 190 to 210 59.1 39.9 25.8 7.01 6.11 210 to 230 112 49.9 24.6 6.56 5.75 230 to 250 33.8 16.9 7.78 5.81 5.14 250 to 340.16 5.37 0.913 1.46 1.52 1.00 Maximum Relative Error 1.9% 1.2% 1.8% 1.1% 1.1%

(1) Accounts for cumulative effect of tolerances (+.05/-.01 on 1.50 thick inner shell and +/-.12 on 8.00 thick gamma shell) of steel thicknesses and tolerances (+/-.12 on 4.50) of resin on side of the cask.

(2) Contact dose rates at axial coordinates less than -187 or greater than 220 cm (in bolded cells of the table) are inside of impact limiters.

(3) The maximum 2 m dose rate at the axis of the trunnion (z=170 cm) is 7.33 mrem/hour.

(4) The maximum 2 m dose rate at the axis of the trunnion (z=170 cm) is 6.75 mrem/hour.

Page 5-76

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 5-20 Response Functions Employed for Fuel Qualification Response Function Elower Eupper (mrem/hour per (MeV) (MeV) particle) 0.01 0.05 0 0.05 0.10 0 0.10 0.20 0 0.20 0.30 0 0.30 0.40 0 0.40 0.60 8.07E-18 0.60 0.80 7.81E-16 0.80 1.00 5.36E-15 1.00 1.33 4.01E-14 1.33 1.66 1.83E-13 1.66 2.00 5.70E-13 2.00 2.50 1.56E-12 2.50 3.00 3.41E-12 3.00 4.00 7.28E-12 4.00 5.00 0 5.00 6.50 0 6.50 8.00 0 8.00 10.00 0 Neutron 1.93E-08 N,Gamma 5.84E-09 Page 5-77

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 5-1 Cask Shielding Configuration Page 5-78

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Assumed Bounding Profile Ref. 3 NUREG Curve Maximum Profile 1.200 1.100 1.000 relative burnup 0.900 0.800 0.700 0.600 0.500 0.400 0.000 0.200 0.400 0.600 0.800 1.000 fractional axial height Figure 5-2 Axial Burnup Profile for Design Basis Fuel Page 5-79

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Dimensions are shown in inches. Numbers in parenthesis correspond to surface numbers in MCNP model.

Figure 5-3 Side View of TN-40 Transport MCNP Model Page 5-80

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 (862) 11.25 (861) 12.01 z = 69.92 (863) 5 3.13 P1 3.13 (866) 4.5 (865) 8.88 (864) 9.5 z = -77.83 Enlarged view of trunnion area. Dimensions are shown in inches. Numbers in parenthesis correspond to surface numbers in MCNP model.

Figure 5-4 Detail Views of TN-40 Transport MCNP Model Page 5-81

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 5-5 Plan View of TN-40 Transport MCNP Model Basket Structure Page 5-82

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Poison Plate Gap SS Compartment Active Fuel Al Plate Unit Cell Al Rail Cask Body Outer Half Cell Figure 5-6 Details of Lattice Unit Cell and Rails/Outer SST Page 5-83

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 5-7 DELETED Figure 5-8 DELETED Page 5-84

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 6 CRITICALITY EVALUATION TABLE OF CONTENTS Discussion and Results........................................................................... 6-1 Package Fuel Loading ............................................................................ 6-3 Model Specification ................................................................................. 6-4 Description of Calculational Model ............................................... 6-4 Package Regional Densities ...................................................... 6-13 Criticality Calculations ........................................................................... 6-19 Calculational Method.................................................................. 6-20 Scoping Calculations and Results .............................................. 6-23 Criticality Calculations and Results ............................................ 6-36 Sensitivity of keff to Number of Isotopes for Burnup Credit ......... 6-38 Sensitivity of keff to Am-241 ........................................................ 6-39 Sensitivity of keff to Gd-155......................................................... 6-39 Sensitivity of keff to BE Ratio ...................................................... 6-39 Sensitivity of keff to Specific Power ............................................. 6-40 Summary of Sensitivity Evaluations ........................................... 6-40 Critical Benchmark Experiments ........................................................... 6-40 Benchmark Experiments and Applicability ................................. 6-42 Results of the Benchmark Calculations ...................................... 6-44 References ........................................................................................... 6-46 Input File Listing .................................................................................... 6-48 SAS2H Input Deck for Design Basis Fuel Assembly -

Zone 8 ........................................................................................ 6-48 CSAS25 Input Deck for Design Basis Criticality Case ............... 6-50 Page 6-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF TABLES Table 6-1 Minimum Burnup as a Function of Enrichment ..................................... 6-64 Table 6-2 Parameters for PWR Assemblies For Shipment ................................... 6-65 Table 6-3 Required Fuel Assembly and Reactor Parameters for SAS2H Models .................................................................................................. 6-65 Table 6-4 Axial Burnup Profiles from Reference [10] ............................................ 6-66 Table 6-5 Modified Axial Burnup Profiles Used for SAS2H Depletion Analysis ................................................................................................ 6-67 Table 6-6 BPRA/RCCA Design Parameters for SAS2H Models ........................... 6-68 Table 6-7 Burnup Dependent Horizontal Burnup Gradients .................................. 6-69 Table 6-8 Basket and Cask Design Dimensions for the CSAS25 Models ............. 6-70 Table 6-9 Description of the KENO Model ............................................................ 6-71 Table 6-10 Correction Factors for SAS2H Isotopic Content .................................... 6-72 Table 6-10a Best-Estimate Correction Factors for SAS2H Isotopic Content ............ 6-83 Table 6-10b Direct Difference Criticality Evaluation Results ..................................... 6-84 Table 6-10c Benchmark Measured Concentrations .................................................. 6-87 Table 6-10d SCALE 4.4 SAS2H Benchmark Results for Actinides and Fission Products ................................................................................... 6-93 Table 6-10e Fission Product and Actinide Correction Factors for TN-40 .................. 6-94 Table 6-10f SAS2H Benchmark Case Physical Parameters .................................... 6-95 Table 6-11 Burned Fuel Isotopic Composition ........................................................ 6-98 Table 6-12 Material Property Data .......................................................................... 6-99 Table 6-13 Material ID IN KENO ........................................................................... 6-100 Table 6-14 Most Reactive Configuration - Fresh Fuel Assumption ...................... 6-101 Table 6-15 Most Reactive Configuration - Burned Fuel ....................................... 6-102 Table 6-16 Results of Burnup Credit Sensitivity Calculations ............................... 6-103 Table 6-16a Burnup Credit Criticality Calculation Results....................................... 6-105 Table 6-16b Horizontal Bias Calculation Results .................................................... 6-107 Table 6-16c Criticality Sensitivity Calculation Results............................................. 6-108 Table 6-16d Moderator Density Variation Results .................................................. 6-109 Table 6-16e DELETED ........................................................................................... 6-109 Table 6-17 Results of the Fuel Assembly Misload Evaluation .............................. 6-110 Table 6-18 Results of the Additional Reactivity Margin Calculations .................... 6-111 Page 6-ii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 6-19 Dancoff Factor Calculation for Density Variations ............................... 6-112 Table 6-20 CSAS25 Results ................................................................................. 6-113 Table 6-20a 142 Experiment Benchmark Comparison SCALE 4.4 vs. SCALE 5.0 ....................................................................................................... 6-118 Table 6-20b CRC Benchmark Results with SCALE 5.0 and the 238 - Group Library ................................................................................................. 6-122 Table 6-20c CRC Benchmark Results with SCALE 5.0 and the 44 - Group Library ................................................................................................. 6-123 Table 6-21 USL-1 Results ..................................................................................... 6-124 Table 6-22 USL Determination for Criticality Analysis........................................... 6-125 Table 6-23 Summary of Criticality Analysis Results .............................................. 6-126 Table 6-24 CRC Crystal River Unit 3 KENO keff with 28-Isotopes ......................... 6-127 Table 6-24a KENO keff with SAS2H 44-Group Depletion Results ........................... 6-128 Table 6-24b Reactivity Worth of Am-241 in CRC experiment ................................. 6-128 Table 6-24c DELETED ........................................................................................... 6-128 Table 6-24d Sensitivity of Specific Power ............................................................... 6-129 Table 6-24e BE Sensitivity ...................................................................................... 6-130 Table 6-24f Correction Factor Sensitivity ............................................................... 6-130 Table 6-25 Comparison of LCE, CRC and TN-40 ................................................. 6-131 Page 6-iii

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 LIST OF FIGURES Figure 6-1 TN-40 Loading Curve (with and without BPRAs) ................................ 6-132 Figure 6-2 Example SAS2H Model ...................................................................... 6-133 Figure 6-3 Fuel Assembly Positions within the Basket ......................................... 6-134 Figure 6-4 Radial Cross Section of the Basket with Centered Fuel Assemblies ......................................................................................... 6-135 Figure 6-5 Radial Cross Section of the Basket with Stainless Steel..................... 6-136 Figure 6-6 Axial Cross Section of the Basket with Cuboid Plugs ......................... 6-137 Figure 6-7 Radial Cross Section of the Basket with Inward Fuel Assemblies ...... 6-138 Figure 6-8 TN-40 KENO Straight Model for Horizontal Burnup Gradient ........... 6-139 Figure 6-8a KENO Plot of the Horizontal Profile - Configuration 1 ........................ 6-140 Figure 6-8b KENO Plot of the Horizontal Profile - Configuration 2 ........................ 6-141 Figure 6-8c KENO Plot of the Horizontal Profile - Configuration 3 ........................ 6-142 Figure 6-8d KENO Plot of the Horizontal Profile - Configuration 4 ........................ 6-143 Figure 6-8e KENO Plot of the Horizontal Profile - Configuration 5 ........................ 6-144 Figure 6-9 TN-40 KENO Model with Internal Moderator between Poison Plates .................................................................................................. 6-145 Figure 6-10 TN-40 KENO Model for Cask A Loading Configuration ...................... 6-146 Figure 6-11 TN-40 KENO Model for Cask B Loading Configuration ...................... 6-147 Figure 6-12 TN-40 KENO Model for Cask C Loading Configuration ...................... 6-148 Figure 6-13 Trending Curves for Direct Difference Evaluation ............................... 6-149 Figure 6-14 Actinide keff Distributions as a Function of BE Ratio ......................... 6-151 Figure 6-15 Fission Product Isotopic Comparisons as a Function of BE Ratio ...... 6-152 Figure 6-16 Gd-155 Isotopic Comparisons as a Function of BE Ratio ................... 6-153 Figure 6-17 Fuel Assembly Inventory at Prairie Island ........................................... 6-154 Page 6-iv

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

6.0 CRITICALITY EVALUATION

The TN-40 cask, as transported, will provide criticality control to meet the criticality performance requirements specified in Sections 71.55 and 71.59 of 10 CFR Part 71 [2].

The criticality control design ensures that the effective multiplication factor (keff) of the contained fuel is no greater than an Upper Subcritical Limit (USL) for the most reactive configuration. The USL includes a confidence band with an administrative safety margin of 0.05. The design has a Criticality Safety Index (CSI, given in 10 CFR 71.59(b) as CSI = 50/N) of 0 because N is infinity (). The number N is based on all of the following conditions being satisfied, assuming packages are stacked together in any arrangement and with close full reflection on all sides of the stack by water:

1. Five times N undamaged packages with nothing between the packages are subcritical;

2. Two times N damaged packages, if each package is subjected to the tests specified in 10 CFR Part 71.73 (HAC) is subcritical with optimum interspersed hydrogenous moderation; and

3. The value of N cannot be less than 0.5.

Discussion and Results The TN-40 basket uses fixed neutron poison plates (or poison plates) for criticality control. The stainless steel basket consists of tubular fuel compartments held together via discrete axial welds forming a 40-compartment basket. The assembly of fuel compartments is connected to aluminum plates at the basket periphery. The aluminum plates provide the circular perimeter geometry that fits the basket inside the cask inner shell and provide for efficient heat transfer from the basket to the cask body. The poison plates are confined between the tubular fuel compartments and the grouped compartments as shown in the various figures in Chapter 1.

The TN-40 cask is shown to be subcritical for an infinite array of flooded undamaged casks and for an infinite array of damaged casks after being subjected to Hypothetical Accident Conditions (HAC) events. The design has a CSI of 0 as N is equal to . A CSI of 0 (less than 50) ensures that, per 10 CFR Part 71.59 (c)(1), the package may be shipped by a carrier in a nonexclusive conveyance, from a criticality requirements point of view.

Page 6-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The calculations performed to confirm the subcriticality requirements listed above utilize a credit for the fuel assembly burnup or burnup credit. Taking burnup credit requires a different analytical approach for criticality analysis than is used in traditional analysis with a fresh fuel assumption. For fresh fuel, the only key fuel parameters to be taken into account in the analyses are the initial enrichment and the most reactive fuel configuration. The analysis of burned fuel must include consideration of the most reactive assembly as a function of burnup, end effects (underburned fuel at the ends),

reactor operating history, fuel composition, initial enrichment and cooling time.

Therefore, additional calculations and codes are required for burned fuel to determine the isotopic composition of the burned fuel as a function of fuel design, initial enrichment, burnup, and cooling time using an assumed bounding reactor operating history. In addition, the benchmarking method to determine code biases is different.

For the criticality code, additional benchmarks are required to account for the burned fuel composition. An additional bias (correction factors) for the depletion code, which determines the isotopic composition of the burned fuel, must also be addressed in the evaluation. Further, an additional bias for the criticality analysis code that accounts for the reactivity effect of the burned fuel must also be addressed in this evaluation.

The depletion calculations determine the isotopic composition of the burned fuel with the SAS2H control module of SCALE-4.4 [1]. The bias due to the fission product absorber isotopic composition of the fuel is accounted for by adjusting the calculated isotopic content based on comparison with available measure isotopic data of burned fuel from a variety of reactors and operating histories. The correction factors are based on the SAS2H benchmarks of measured data. The bias due to the actinide fissile isotopic composition of the fuel is accounted for by a direct difference method [13]. The two biases are documented in Section 6.3.2. The criticality calculations determine keff with the CSAS25 control module of SCALE-4.4. The bias due to the criticality code with the additional benchmark data to account for the composition of the burned fuel is also included in Section 6.5.2.

A series of scoping calculations with the CSAS25 control module of SCALE-4.4 determines the most reactive configuration for the basket and assembly location. It also evaluates all of the eligible fuel assembly designs allowed for transport in the TN-40 package and determines that the Westinghouse 14x14 Standard fuel assembly is the design basis fuel assembly because it is the most reactive fuel assembly authorized for transport. The results of those scoping calculations are documented in Section 6.4.2.

With the CSAS25 control module of SCALE-4.4, a series of criticality calculations analyzes a set of bounding configurations that are determined based on the most reactive configuration. Those bounding configurations are used to evaluate the horizontal bias, the TN-40 loading curve and fuel misload evaluations. They are characterized by initial enrichment, assembly average burnup, and a minimum cooling time in addition to the geometric arrangement in the most reactive configuration. The results of those criticality calculations are documented in Section 6.4.3. A series of sensitivity calculations has also been performed with the CSAS25 control module of SCALE 4-4. It addresses some specific issues associated with fission products and actinides. The results of those sensitivity calculations are documented in Section 6.4.4.

Page 6-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The main results of the criticality calculations are shown in Table 6-1 and Figure 6-1.

The minimum assembly average burnup as a function of initial enrichment required to ensure subcriticality is shown in Table 6-1 for fuel assemblies irradiated with and without BPRAs. A polynomial function is utilized to fit this data so that the required assembly burnup for all assembly enrichments that lie in between the minimum and maximum can be calculated. A separate curve is fit for fuel assemblies with and without BPRAs. A burnup curve based on the results in Table 6-1 is shown in Figure 6-1. The acceptable region in the curve pertains to those fuel assemblies with a burnup-enrichment combination AND a minimum cooling time greater than 30 years, that are eligible for transportation in the TN-40 cask.

In conclusion, the results of the criticality calculations and the sensitivity calculations demonstrate that the maximum keff, including statistical uncertainty, is less than the USL determined from a statistical analysis of benchmark criticality experiments and includes an allowance for a burned fuel. The statistical analysis procedure includes a 95%

confidence band with an administrative safety margin of 0.05.

Package Fuel Loading The TN-40 Cask is capable of transporting 40 undamaged Westinghouse 14x14 (WE

14) class of PWR fuel assemblies irradiated at the Prairie Island Nuclear Generating Plant with or without Non Fuel Assembly Hardware (NFAH). Burnable Poison Rod Assemblies (BPRAs) and rod cluster control assemblies (RCCAs or Control Rods) are the only NFAH that are discussed in this evaluation since they bound all other NFAH.

Each BPRA typically consists of 4, 8, 12, or 16 burnable poison (BP or discrete BP) rods. Each RCCA consists of 16 control rods or fingers. The fuel assemblies considered as authorized contents (as discussed in Chapter 1) include those listed in Table 6-2. Table 6-2 also lists the fuel parameters for the PWR fuel assemblies. The design basis fuel assembly for the TN-40 cask criticality analysis was determined to be the WE 14 Standard fuel assembly.

Page 6-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 6-4 through 6-45 Withheld Pursuant to 10 CFR 2.390 Page 6-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. Oak Ridge National Laboratory, RSIC Computer Code Collection, SCALE: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluations for Workstations and Personal Computers, NUREG/CR-0200, Revision 6, ORNL/NUREG/CSD-2/V2/R6.

2. 10 CFR 71, Packaging and Transportation of Radioactive Materials.

3. M. D. DeHart SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1 - Summary, Oak Ridge National Laboratory, March 1995, ORNL/TM-12294/V1.

4. U.S. Nuclear Regulatory Commission, Criticality Benchmark Guide for Light-Water-Reactor fuel in Transportation and Storage Packages, NUREG/CR-6361, Published March 1997, ORNL/TM-13211.

5. U.S. Nuclear Regulatory Commission, Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, NUREG/CR-5661, Published April 1997, ORNL/TM-11936.

6. Oak Ridge National Laboratory, An Extension of the Validation of SCALE (SAS2H)

Isotopic Predictions for PWR Spent Fuel, ORNL/TM-13317, Published September 1996.

7. U.S. Nuclear Regulatory Commission, Parametric Study of the Effect of Control Rods for PWR Burnup Credit, NUREG/CR-6759, Published February 2002, ORNL/TM-2001/69.

8. U.S. Nuclear Regulatory Commission, Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit, NUREG/CR-6760, Published March 2002, ORNL/TM-2000-329.

9. U.S. Nuclear Regulatory Commission, Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit, NUREG/CR-6761, Published March 2002, ORNL/TM-2000/373.

10. U.S. Nuclear Regulatory Commission, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, NUREG/CR-6801, Published March 2003, ORNL/TM-2001/273.

11. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages, DOE/RW-0472, Revision 2.

12. U.S. Nuclear Regulatory Commission, Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs, NUREG/CR-6800, Published March 2003, ORNL/TM-2002/6.

Page 6-46

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

13. U.S. Nuclear Regulatory Commission, Strategies for Application of Isoptic Uncertainties in Burnup-Credit, NUREG/CR-6811, Published June 2003, ORNL/TM-2001/257.

14. Oak Ridge National Laboratory, Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analysis, ORNL/TM-12667, Published March 1995.

15. CAL-UDC-NU-000011 Rev A, Three Mile Island Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations, Office of Civilian Radioactive Waste Management, U.S. Department of Energy, April 2002.

16. U.S. Nuclear Regulatory Commission, Isotopic Analysis of High-Burnup PWR Spent Fuel Samples from the Takahama-3 Reactor, NUREG/CR-6798, Published January 2003, ORNL/TM-2001/259.

17. Radulescu G, Mueller D. E. and J. C. Wagner, Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit, Oak Ridge National Laboratory, January 2008, ORNL/TM-2006-87, NUREG/CR-6951.

18. Criticality Model, CAL-DS0-NU-0000003 REV 00A, Bechtel SAIC Company, Las Vegas, Nevada, 2004.

19. S. M. Bowman, W.C. Jordan, J. F. Mincey, C.V. Parks, and L. M. Petrie, Experience with the SCALE Criticality Safety Cross-Section Libraries, Oak Ridge National Laboratory, NUREG/CR-6686, Published October 2000, ORNL/TM-1999/322.

20. C. V. Parks, M. D. DeHart, and J. C. Wagner, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, NUREG/CR-6665, ORNL/TM-1999/303, Oak Ridge National Laboratory, February 2000.

Page 6-47

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 6-48 through 6-123 Withheld Pursuant to 10 CFR 2.390 Page 6-48

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 6-21 USL-1 Results Range of Formula for USL-1 Parameter applicability (0.05 keff Margin)

U Enrichment 0.9406 + ( 9.7267E-04)*X (X < 3.7186 )

(wt% U-235) 2.35 - 5.74 0.9442 (X >= 3.7186 )

(fresh fuel experiments)

U Enrichment 0.9389 + ( 7.3145E-04)*X (X < 4.5697 )

(wt% U-235) 2.35 - 5.74 0.9422 (X >= 4.5697 )

(all experiments)

Pu Enrichment 2.0 - 6.6 0.9424 (wt% Pu)

Fuel Rod Pitch (cm) 0.9351 + ( 5.2323E-03)*X (X < 1.7753 )

1.10 - 2.64 (fresh fuel experiments) 0.9443 (X >= 1.7753 )

Fuel Rod Pitch (cm) 0.9330 + ( 5.2634E-03)*X (X < 1.8164 )

1.10 - 2.64 (all experiments) 0.9425 (X >= 1.8164 )

Water/Fuel Volume Ratio 0.9416 + ( 7.3797E-04)*X (X < 2.1094 )

0.38 - 10.8 (fresh fuel experiments) 0.9431 (X >= 2.1094 )

Water/Fuel Volume Ratio 0.9387 + ( 9.2625E-04)*X (X < 2.6949 )

0.38 - 10.8 (all experiments) 0.9412 (X >= 2.6949 )

Assembly Separation (cm) 0.9410 + ( 4.9375E-04)*X (X < 6.9867 )

1.64 - 20.78 (fresh fuel experiments) 0.9444 (X >= 6.9867 )

Energy of the Average Lethargy for Fission (EALF) 0.0083 - 1.39 0.9434 (Entire Range)

(fresh fuel experiments)

Energy of the Average 0.9421 (X <= 0.2069 )

Lethargy for Fission (EALF) 0.0083 - 1.39 0.9427 + (-3.0241E-03)*X (X > 0.2069 )

(all experiments)

Page 6-124

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Table 6-22 USL Determination for Criticality Analysis Value from Limiting Parameter Bounding USL WE 14x14 Analysis Enrichment (wt. % U-235) 2.25 (minimum)(1) 0.9406 (all benchmarks)

Enrichment (wt. % U-235) 2.00 (minimum)(1) 0.9425 (fresh fuel benchmarks)

Enrichment (wt. % Pu) Not relevant since there is 0.9424 (fresh fuel benchmarks) no variation in the USL Pin Pitch (cm) 1.412 0.9403 (all benchmarks)

Pin Pitch (cm) 1.412 0.9424 (fresh fuel benchmarks)

Water to Fuel Volume Ratio 1.610 (2) 0.9402 (all benchmarks)

Water to Fuel Volume Ratio 1.610 (2) 0.9427 (fresh fuel benchmarks)

Assembly Separation (cm) 1.92 (3) 0.9419 (fresh fuel benchmarks)

Energy of the Average Lethargy for Fission (EALF) 0.35 (4) 0.9416 (all benchmarks)

Energy of the Average Lethargy for Fission (EALF) 0.35 (4) 0.9434 (fresh fuel benchmarks)

1) Extrapolation of the USL-1 formula is performed at this enrichment to determine the minimum USL since the keff data showed no trending with enrichment.

2) The water to fuel volume ratio is calculated using 179 rods.

3) Separation Distance = 2*(0.09") + 2*(0.250) + 0.075" = 0.755" ~ 1.92 cm, calculated with nominal dimensions for the stainless steel in the fuel compartment, nominal boral and aluminum plate width and inward fuel assembly positioning.

4) Examination of the results shows that the value is between 0.25 and 0.35 and hence a conservative value that produces the minimum USL was chosen even though there is no variation in the USL with fresh fuel experiments.

Page 6-125

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 6-126 through 6-131 Withheld Pursuant to 10 CFR 2.390 Page 6-126

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 POLYNOMIAL FITS FOR LOADING CURVES 38000 36000 34000 32000 30000 y = -1259.8x2 + 20242x - 23617 R2 = 0.9956 28000 BURNUP (MWD/MTU) 26000 24000 22000 20000 y = - 366.95x2 + 14770x - 17200 R2 = 0.9966 18000 BPRA-CURVE 16000 NOBP-CURVE 14000 12000 10000 2.00 2.20 2.40 2.60 2.80 3.00 3.20 3.40 3.60 3.80 4.00 ENRICHMENT (WT. % U-235)

Note: The ACCEPTABLE region for the BPRA-CURVE always lies above the curve and corresponds to a burnup that is greater than or equal to the burnup loading curve. The ACCEPTABLE region for the NOBP-CURVE always lies above the curve and corresponds to a burnup that is greater than or equal to the burnup loading curve.

Figure 6-1 TN-40 Loading Curve (with and without BPRAs)

Page 6-132

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 6-2 Example SAS2H Model Page 6-133

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Proprietary Information on Pages 6-134 through 6-153 Withheld Pursuant to 10 CFR 2.390 Page 6-134

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure RAI 6-5 Prairie Island Fuel vs Loading Curve 55000 50000 45000 Burnup (MWd/MTU) 40000 35000 30000 25000 20000 15000 10000 5000 0

2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 Maximum Initial Enrichment (% )

Note: The maximum enrichment for the Westinghouse 14x14 Standard fuel type is 3.4 wt. % U-235 Figure 6-17 Fuel Assembly Inventory at Prairie Island Page 6-154

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 APPENDIX 6A ASSIGNED BURNUP LOADING VALUE TABLE OF CONTENTS Loading the Wrong Fuel Assembly ............................................6A-1 6A.1.1 Summary of Preventive Measures ................................6A-2 Calculating a Burnup Value Higher Than Actual ........................6A-2 6A.2.1 Summary of Preventive Measures ................................6A-3 Wrong Burnup Value Assigned to a Fuel Assembly ...................6A-3 6A.3.1 Summary of Preventive Measures ................................6A-5 Conclusion .................................................................................6A-5 LIST OF FIGURES Figure 6A-1 Burnup Calculation Flow Path....................................................... 6A-6 Page 6A-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The following outlines the administrative procedures that will be established to provide protection against the loading a fuel assembly with a burnup value less than that required by the loading curve. The outline is broken down into the three most likely ways that a misloading could occur 1) loading the wrong fuel assembly, 2) calculating a burnup value higher than actual, and 3) wrong burnup value assigned to a fuel assembly.

Loading the Wrong Fuel Assembly One possible misloading that could result in a fuel assembly being placed into a cask that does not satisfy the burnup requirements is that a fuel assembly that is not intended to be loaded is actually loaded into a cask. This could happen in one of two ways. The first way being a fuel assembly was stored in the wrong location within the spent fuel pool. The second way being the operator removed the fuel assembly from the wrong spent fuel location.

For example:

1st misloadingPlan calls for placing fuel assembly E-21 stored in spent fuel pool location S.H43 into cask. However, fuel assembly Q-09 is actually being stored in Location S.H43.

2nd misloadingPlan calls for placing fuel assembly E-21 stored in spent fuel pool location S.H43 into cask. However, the operator inadvertently loads fuel assembly Q-09 from spent fuel pool location S.H44 into the cask.

The criticality risk associated with the first type of misloading will be reduced by verifying the identity of each fuel assembly at some time prior to loading the fuel into the cask.

This will provide some assurance that when the cask is loaded that the correct fuel assemblies are in the designated spent fuel pool locations.

The criticality risk associated with both types of misloadings will be reduced by requiring the boron concentration in the spent fuel pool water and any water to be added to the cask, during loading operations, to be borated to 1800 ppm. Since the above risks are identical for the loading a TN40 storage cask, the required boron concentration is based on the required concentration contained in the Prairie Island Independent Spent Fuel Storage Installation Technical Specifications (Docket Number 72-10 and License Number SNM-2506) for the TN-40 Storage casks.

To ensure that the cask contains the fuel that satisfies the loading requirements, the identity of the fuel that has been loaded will be verified prior to closing the cask.

These actions will reduce the criticality risk associated with loading a wrong fuel assembly into a TN40 transport cask.

Page 6A-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 6A.1.1 Summary of Preventive Measures

1) Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of placing the first fuel assembly into the cask. Verify that dissolved boron concentration in the spent fuel pool water and water to be added to the cask cavity is greater than 1800 ppm.

2) Once prior to inserting into cask and once prior to closure of cask. Verify the identity of each fuel assembly. This verification shall be performed by two independent individuals.

Calculating a Burnup Value Higher Than Actual The second way to have a misloading is for the actual burnup to be less than the calculated burnup due to uncertainties in the calculation.

For example:

The records indicate that fuel assembly J-52 (which had an initial enrichment of 3.82% w/o U-235) has a burnup value of 31,639 MWD/MTU. At an enrichment of 3.82%, the allowable burnup limit would be 30,660 MWD/MTU. Thus the records would indicate that assembly J-52 satisfies the loading requirement. However, due to uncertainties in the calculation of burnup, the actual burnup of J-52 is 30,500 MWD/MTU.

Figure 6A-1 provides a simplified flow path of information and computer codes used to calculate an individual fuel assemblys burnup. Note that this simplified flow path does not show all input decks that are needed to run the codes but only the key information related to calculating a fuel assemblys burnup.

The process starts by collecting hourly power history from either the calibrated Nuclear Power Range instruments or directly from the online calorimetric for a given period of time, typically one month. This data along with the total core metric ton uranium loading is input to the FOLLOW code that then calculates the total core burnup over that period.

Independently of the FOLLOW run, measured incore flux data is processed by the INCORE or DETECTOR codes to determine relative measured power distribution. This power distribution is used to demonstrate compliance with the Peaking Factor Limits provided in the sites Technical Specifications as well as providing individual fuel assembly burnup rate information.

The TOTE or BURNUP codes combine the individual fuel assembly burnup rate information with the total core burnup for the period to determine the fuel assemblys burnup for that period. This is then added to the fuel assemblys previous burnup to obtain an accumulated burnup.

The sources of uncertainty in this calculation process are best understood by looking at the units of burnup, i.e., MWD/MTU.

Page 6A-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The amount of uranium loaded into a fuel assembly is known to a high degree of accuracy as listed by the reporting to the gram on the DOE/NRC Form 741 Nuclear Material Transaction Report for fuel assemblies containing approximately 400 kilograms of uranium. Therefore the uncertainty in the loading of an assembly is negligible.

The uncertainty associated with the energy release, i.e., MWD, comes from two sources, the first being the power uncertainty in the hourly power history data and the second is in the measured relative power distribution.

The actual power uncertainty over an operating cycle is typically on the order of 1%.

This is supported by the historically good agreement between the end of hot full power cycle exposures and the predicted values at Prairie Island. As stated earlier the measured relative power distribution from the INCORE and DETECTOR codes is used to demonstrate compliance with the sites Peaking Factor Technical Specifications. One of these peaking factors, FH, is essentially a radial pin power peaking factor and Technical Specification specified that the measured value be multiplied by 1.04 to account for measurement uncertainties. Since the uncertainty of measuring the power of an individual fuel pin would be larger than measuring the power of an entire fuel assembly, it is reasonable to conclude that the power measurement uncertainty, and thus the burnup rate uncertainty, of an assembly would be less than 4%. It is also important to remember that on average, the uncertainties in the relative burnup rate for a given period would cancel out and such that the average uncertainty would be no more than that associated with the hourly power history uncertainty.

Recognizing that:

1) On average the uncertainty of the relative burnup rates cancel out.

2) The uncertainty of an individual fuel assembly burnup rate is less than 4%.

3) The uncertainty in the hourly power history is on the order of 1%.

4) The uncertainty in the uranium loading is negligible.

It is proposed that the individual fuel assembly burnup from the TOTE or BURNUP code be reduced by a factor of 1.04 to account for the measurement uncertainty in the calculated burnup.

6A.2.1 Summary of Preventive Measures

1) The assigned burnup loading value for each fuel assembly shall be from the TOTE or BURNUP computer codes. The value from these codes shall be reduced by a 1.04 factor to account for burnup uncertainties.

Wrong Burnup Value Assigned to a Fuel Assembly The third way for a misloading to occur is to have the assigned burnup value for at least two assemblies interchanged.

Page 6A-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 For example:

Fuel assembly M-71 has actual discharge exposure of 29,040 MWD/MTU and assembly L-58 has an actual discharge exposure of 32,715 MWD/MTU. Both of these assemblies have an initial U-235 enrichment of 3.82% and thus an allowable burnup limit of 30,660 MWD/MTU. Therefore assembly L-58 satisfies the loading requirement while assembly M-71 does not.

However, the records of these assemblies have become interchanged such that they indicate that L-58 has an exposure of 29,040 MWD/MTU and the records for M-71 indicate 32,715 MWD/MTU and thus indicating that M-71 would satisfy the loading requirements.

There are two possible ways for this type of interchange to occur. The first is that the fuel assembly identities are interchanged in the input decks of the code that calculates the burnup. The second is that the information becomes interchanged somewhere between when it is initially transcribed from the output of the computer runs to the final documentation that the fuel assembly satisfies the loading requirements.

The risk associated with interchanging identities in the input decks is small because of the care taken in preparing those decks for a given cycle of operation. Although the burnup calculations are performed monthly, the base input file(s) need only be setup at the beginning of the cycle. These files would then be used each month changing only those inputs reflecting that months total core burnup. Since the decks are setup for each individual cycle, it is not plausible to interchange assemblies from different cycles of operation. Most identity or location interchanges would become self evident during the monthly runs of the code. For example if the identity of a twice burned assemble were to have been interchanged with that of a fresh or once burned assembly it would typically be very obvious in the output, i.e., the affected assemblys burnup would not be consistent with the remainder of the assemblies of the same batch or region. If two assemblies of a given batch (and thus have the same enrichment) were to have their identities interchanged, the risk of loading an assembly that does not meet the burnup requirement is small since for a given enrichment all but a very few fuel assemblies satisfy the allowable burnup limit.

The risk associated with the records of two fuel assemblies becoming interchanged somewhere between recording the data from the TOTE or BURNUP output and the document showing that the fuel assembly satisfies the allowable burnup limit will be reduced by requiring that the assigned burnup loading value for each fuel assembly be obtained from a source (e.g., formal calculation or data base) that is controlled by the sites Quality Assurance Program. The data within this source shall be traceable to the TOTE or BURNUP output corresponding to when the fuel assembly was discharged from the reactor for the final time.

Page 6A-4

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 6A.3.1 Summary of Preventive Measures

1) Once prior to placing the first fuel assembly into the cask. Verify each fuel assembly to be loaded satisfies the loading requirements. This verification shall be performed by two independent individuals.

2) The assigned burnup loading value for each fuel assembly shall be obtained from a source controlled by the sites QA program and traceable to the TOTE or BURNUP output corresponding to when the fuel assembly was discharged from the reactor for the final time.

Conclusion The above administrative controls will provide the necessary protection against the loading a fuel assembly with a burnup value less than that required by the loading curve.

Page 6A-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 6A-1 Burnup Calculation Flow Path Page 6A-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Chapter 7 Operating Procedures TABLE OF CONTENTS 7.0 OPERATING PROCEDURES ........................................................................... 7-1 Package Loading .................................................................................... 7-1 Preparation for Loading .............................................................. 7-1 Loading of Contents .................................................................... 7-2 Preparation for Transport ............................................................ 7-2 Package Unloading ................................................................................. 7-4 Receipt of Package from Carrier................................................. 7-4 Preparation for Unloading ........................................................... 7-5 Removal of Contents .................................................................. 7-6 Preparation of Empty Package for Transport .......................................... 7-8 References ........................................................................................... 7-10 LIST OF FIGURES Figure 7-1 Torqueing Patterns ............................................................................... 7-11 Figure 7-2 Typical Setup for Filling Cask with Water.............................................. 7-12 Page 7-i Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 7.0 OPERATING PROCEDURES This chapter contains the TN-40 loading and unloading procedures that are intended to show the general approach to cask operational activities. A separate Operation Manual (OM) will be prepared for the TN-40 transport package to describe the operational steps in greater detail. The OM, along with the information in this chapter, will be used to prepare the site-specific procedures that will address the particular operational considerations related to the TN-40.

The procedures in this chapter are intended to show the types of operations that will be performed and are not intended to be limiting. Site specific conditions and requirements may necessitate the use of different equipment and ordering of steps to accomplish the same objectives or acceptance criteria which must be met to ensure the integrity of the package. Deviations to the provided procedures are acceptable if justified by the applicable Licensee or Certificate Holder in their QA program to maintain equal or better package effectiveness and continued compliance with the applicable 10 CFR Part 71 requirements.

Package Loading Preparation for Loading NOTE:

The steps in this procedure assume a TN-40 in use for storage under a 10 CFR Part 72 license. Preparation for the loading section of this procedure includes only steps to verify that the cask packaging complies with the requirements in Chapter 1, Appendix 1.4.1, and Chapter 8.

1. Review the fabrication, maintenance, and design change control records for each cask to verify that the configuration of the cask as used for storage complies with the conditions for approval for use as a transportation package.

2. Review the fabrication records for the cask to verify that all required acceptance testing and inspections have been performed.

3. Review the maintenance records to verify that all required periodic inspections and tests have been performed.

Page 7-1 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Loading of Contents NOTE:

The steps in this procedure assume a TN-40 in use for storage under a 10 CFR Part 72 license. Loading of contents section of this procedure includes only steps to verify that the cask contents comply with the requirements in Chapter 1.

1. Review the loading records to verify each fuel assembly loaded for storage satisfies the requirements listed in CoC No. 9313. This verification shall be performed by two independent individuals. The burnup value shall be reduced by a 1.04 factor to account for burnup uncertainties.

2. Review the loading records to verify the cask was vacuum dried and backfilled with helium during loading for storage.

3. Review the maintenance records of the cask for situations where air may have leaked into the cask while it was in its storage configuration on the storage pad. If air has leaked into the cask while it was in its storage configuration, perform an evaluation prior to transportation of the fuel cladding for potential rod splitting due to exposure to an oxidizing atmosphere using the methodology given in NUREG-2215, Appendix 8C

[3].

Preparation for Transport NOTE:

All threaded fasteners that are to be installed during these operations shall be cleaned and visually inspected for damage prior to installation. All bolt threads shall be coated with Loctite N-5000, nuclear-grade Neolube, or equivalent.

1. Remove the protective cover.

2. Remove the overpressure protection (OP) system, as follows:

a) Disconnect the OP system from the monitoring panel.

b) Depressurize the OP tank and disconnect the tubing at the protective cover.

c) Remove the OP tank assembly, including the OP port cover and the top neutron shield.

3. Confirm that the lid, vent cover, and drain cover bolts are torqued to requirements on Drawing 10421-71-1 using the sequence shown in Figure 7-1.

4. Verify that the lid bolts are the material specified on Drawing 10421-71-1.

If lid bolts require replacement, then the bolts shall be replaced one at a time. This operation will take place indoors.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

5. Visually inspect accessible surfaces of the cask for evidence of cracks, corrosion, dents, or other degradation.

6. Verify that radiation and contamination levels are in accordance with the site requirements to minimize worker exposure. Also check surface contamination levels to verify that they meet the requirements of 49 CFR 173.443 [8] and 10 CFR 71.87 [6].

NOTE:

A periodic leakage rate test, as described in Section 8.2.2, shall be performed within 12 months prior to shipment. The pre-shipment leak test is not required if there has been no disassembly of the lid, vent port cover, or drain port cover, or loosening of any closure bolts (except as covered under Section 7.1.3, Step

4) since the periodic leakage rate test was last performed.

7. Perform a pre-shipment leak test of the lid, vent port cover, and drain port cover seals in accordance with ANSI N14.5 [1]. The combined standard leak rate shall be less than 1.0 x 10-4 ref cm3/s with a test sensitivity of at least 5.0 x 10-5 ref cm3/s, or no detectable leakage when tested to a sensitivity of at least 1.0 x 10-3 ref cm3/s.

8. Replace the storage OP port cover with the transport OP port cover.

CAUTION:

Lower trunnions shall not be used as a lifting attachment point.

9. Place cask on the transport frame, as follows:

a) Move the transport vehicle with transport frame in place into the loading area and prepare to down end the cask.

b) Attach the lift beam to the cask handling crane hook, and then engage the lift beam to the two upper (top) trunnions.

c) Lift the cask and place the rear trunnions on the rear trunnion supports of a separate down ending frame or the transport frame.

d) Rotate the cask from the vertical to the horizontal position. Skip Step 8.e) if the cask is on the transport frame.

e) Using a spreader bar and lifting straps, lift the cask from the down ending frame and lower it onto the transport frame.

10. Install the tie-down straps.

11. Install the impact limiters (ILs), as follows.

a) Prior to installing the ILs, inspect them visually for damage. The ILs may not be used without repair if any wood has been exposed.

b) Install the top IL spacer on the front end (lid end) of the cask.

c) Install the front (top) and the rear (bottom) ILs onto the cask.

Page 7-3 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 d) Thread all the attachment bolts hand-tight and torque them to 60-80 ft-lb.

e) Install all the IL attachment tie-rods between the front and the rear ILs.

f) Render the IL lifting lugs inoperable.

g) Install a security seal on one tie-rod and lock sleeve.

12. Perform a temperature survey to verify compliance with 10 CFR 71.43(g)

[4].

13. Install the personnel barrier.

14. Perform a final radiation survey to assure the cask radiation levels do not exceed 49 CFR 173.441 [7] and 10 CFR 71.47 [5] limits.

15. If the measured dose rate in the normally occupied spaces exceeds 2 mr/hr, the location of the package shall be changed or supplementary shielding added as necessary to reduce the dose to an acceptable level.

Supplementary shielding may be added to the conveyance (e.g., attached to the sides of the trailer or truck cab) to reduce the external radiation levels, but shall not be attached to the package without prior NRC approval. Alternatively, the carrier may implement the radiation dosimetry requirements of 49 CFR 173.441(b)(4) [7] and 10 CFR 71.47(b)(4) [5].

16. Verify the carrier has been provided with written instructions and a list of contacts for notification in case of accident or delays.

17. Verify the carrier has been provided with written instructions for maintenance of the Exclusive Use shipment controls in accordance with 10 CFR 71.47 [5].

18. Apply appropriate U.S. Department of Transportation (DOT) labels and placards in accordance with 49 CFR Part 172 [2].

19. Prepare the final shipping documentation.

20. Release the loaded cask for shipment.

Package Unloading Receipt of Package from Carrier

1. Upon arrival of the loaded cask, perform a receipt inspection of the cask to check for any damage or irregularities. Verify that the security seal is intact, and perform a radiation survey.

2. Remove the personnel barrier.

3. Remove the ILs, as follows:

a) Remove the security seal, tie-rods, and the associated hardware.

b) Remove any features used to render the IL lifting lug holes inoperable.

c) Remove the IL attachment bolts.

Page 7-4 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 d) Remove the front and rear ILs.

e) Remove the top IL spacer.

4. Remove the tie down straps.

5. Verify that the cask surface removable contamination levels meet the requirements of 49 CFR 173.443 [8].

6. Perform a radiation survey of the cask to verify compliance with 10 CFR 71.47 [5].

CAUTION:

Lower trunnions shall not be used as a lifting attachment point.

7. Remove the cask from the transport frame, as follows:

a) Move the transport vehicle into the unloading area and prepare to upend the cask. If upending the cask on the transport frame, then go to Step 7.c).

b) Using a spreader bar and lift slings, lift the cask from the transport vehicle and place it on an upending frame.

c) Attach the lift beam to the cask handling crane hook, and then engage the lift beam to the two upper (top) trunnions.

d) Rotate the cask slowly from the horizontal to the vertical position.

e) Lift the cask, and place it in a designated area.

f) Disengage the lift beam from the cask, and remove from the area.

8. If necessary, clean the external surfaces of the cask to get rid of the road dirt.

Note:

If the cask is being placed into storage, skip Sections 7.2.2 and 7.2.3.

Follow procedures for transfer to site storage in compliance with the 10 CFR Part 72 license.

Preparation for Unloading

1. Remove the neutron shield pressure relief valve, and install the plug in the neutron shield vent hole.

2. Remove the OP port cover.

3. Vent the cask cavity gas, as follows:

a) Remove the vent port cover.

b) Collect a cavity gas sample through the vent port quick-disconnect coupling.

Page 7-5 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 c) Analyze the gas sample for radioactive material, and add necessary precautions based on the cavity gas sample results.

NOTE:

If degraded fuel is suspected, additional measures, appropriate for the specific conditions, are to be planned, reviewed, and approved by the appropriate site personnel, as well as implemented to minimize worker exposures and radiological releases to the environment. These additional measures may include provision of filters, as well as respiratory protection and other methods to control releases and exposures to as low as reasonably achievable (ALARA).

d) In accordance with the site requirements, vent the cavity gas through the vent port until atmospheric pressure is reached.

4. Remove the vent port quick-disconnect and drain port cover. Attach a vent port adapter and the drain port quick-disconnect, if used.

5. Attach the lift beam to the two upper trunnions of the cask. Attach the lid lifting equipment.

CAUTION:

Steam may be present inside the cask. Ensure that appropriate measures are taken to prevent steam burns and radiation exposures. Both fill and vent lines should be designed for steam at a minimum of 100 psig.

6. Attach the vent line to the vent port and attach the water fill line to the drain port.

NOTE:

If the maximum lift weight is not exceeded, the cask may be filled with pool water before lowering the cask into the pool or while the cask is partially submerged in the spent fuel pool.

7. Lower the cask into the spent fuel pool cask pit. Lower the cask until the top surface is just above the water level.

Removal of Contents

1. Begin pumping pool or demineralized water into the cask through the drain port, at a minimum rate of 1 gpm, while continuously monitoring the exit pressure (see Figure 7-2). If the pressure gauge reading exceeds 55 psig, close the inlet valve until the pressure falls below 50 psig. Re-flooding may then be resumed. Continue pumping the water until the water level in the cask has reached the active fuel length.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23

2. The flow rate can be gradually increased, while monitoring the pressure at the outlet. If the pressure gauge reading exceeds 55 psig, close the inlet valve until the pressure falls below 50 psig. Re-flooding may then be resumed.

3. When the cask is full of water, remove the hoses from the drain and vent ports.

4. Loosen and remove all lid bolts.

5. Lower the cask and place it in the cask loading area of the pool.

6. Raise the lift beam from the cask, removing the cask lid.

7. Unload the spent fuel assemblies in accordance with the site procedures.

8. Using the lift beam and lid lifting slings, lower the lid placing it on the cask shell flange, over the two alignment pins. At least one lid penetration must be completely open (both port cover and quick-disconnect fitting removed) prior to installation of the lid.

9. Engage the lift beam on the upper (top) trunnions, and lift the cask out of the pool.

10. Using the drain port in the lid, remove the water from the cask. This may be done either before or after lifting the cask out of the pool. While lifting the cask out of the pool, the exterior of the cask may be rinsed with clean demineralized water to facilitate decontamination. If the pool contains borated water, the effects of adding non-borated water to the pool must be considered in accordance with site procedures.

11. Disconnect the drain line.

12. Move the cask to the decontamination area, and disengage the lift beam.

Page 7-7 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Preparation of Empty Package for Transport NOTE:

All threaded fasteners that are to be installed during these operations shall be cleaned and visually inspected for damage prior to installation. All bolt threads shall be coated with Nuclear Grade Neolube, Loctite N-5000, or equivalent.

1. Verify the cask is empty to the extent practical.

2. Determine the amount and form of residual internal activity inside the empty cask.

3. Inspect and securely close the empty cask.

a) Metal seals may be reused for transport of the empty package.

b) Install all lid bolts and torque to 400 ft-lb. Follow the torqueing sequence shown in Figure 7-1. A circular pattern of torqueing may be used afterwards to eliminate further bolt movement.

4. Remove the plug from the neutron shield vent, and reinstall the pressure relief valve.

5. Evacuate the cask cavity using the vacuum drying system (VDS) to remove any remaining moisture.

6. Isolate the vacuum pump, and backfill the cask cavity with nitrogen.

7. Install the vent and drain port covers, and torque the bolts to requirements on Drawing 10421-71-1 using the sequence shown in Figure 7-1.

CAUTION:

Lower trunnions shall not be used as a lifting attachment point.

8. Place cask on shipping frame, as follows:

a) Move the transport vehicle with transport frame in place into the loading area and prepare to down end the cask.

b) Attach the lift beam to the cask handling crane hook, and then engage the lift beam to the two upper (top) trunnions.

c) Lift the cask, and place the rear trunnions on the rear trunnion supports of a separate down ending frame or the transport frame.

d) Rotate the cask from the vertical to the horizontal position. Skip Step 8.e) if the cask is on the transport frame.

e) Using a spreader bar and lifting straps, lift the cask from the down ending frame and lower it onto the transport frame.

9. Install the tie-down straps.

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 NOTE:

If the ILs are going to be shipped separately, go to Step 11.

10. Install the ILs, as follows:

a) Install top IL spacer on the front end (lid end) of the cask.

b) Install front (top) and the rear (bottom) ILs onto the cask.

c) Thread all lubricated attachment bolts hand-tight and torque to 60-80 ft-lb.

d) Install all IL attachment tie-rods between the front and the rear ILs.

e) Render IL lifting lugs inoperable.

11. Install personnel barrier.

12. Prepare the empty cask for shipment using the package requirements specified in the Hazardous Material Regulations (HMR) [2], which are appropriate for the amount and form of the residual activity and contamination.

13. Release empty cask for shipment.

Page 7-9 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. ANSI N14.5-2014, Leakage Tests on Packages for Shipment of Radioactive Materials.

2. Title 49, Code of Federal Regulations, Subtitle B, Chapter 1, Parts 171 through 180.

3. NUREG-2215, U.S. Nuclear Regulatory Commission, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, Final Report, April 2020.

4. 10 CFR 71.43, General Standards for All Packages.

5. 10 CFR 71.47, External Radiation Standards for All Packages.

6. 10 CFR 71.87, Routine Determinations.

7. 49 CFR 173.441, Radiation Level Limitations and Exclusive Use Provisions.

8. 49 CFR 173.443, Contamination Control.

Page 7-10 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 7-1 Torqueing Patterns Page 7-11 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Figure 7-2 Typical Setup for Filling Cask with Water Page 7-12 Chapter 7 replaced in its entirety in Revision 17

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 CHAPTER 8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM TABLE OF CONTENTS 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ............................. 8-1 Acceptance Tests ................................................................................... 8-1 Visual Inspection .......................................................................... 8-1 Structural and Pressure Tests ...................................................... 8-2 Containment Boundary Leakage Tests ........................................ 8-3 Component Tests ......................................................................... 8-4 Shielding Tests............................................................................. 8-5 Neutron Absorber Tests ............................................................... 8-6 Thermal Acceptance Tests .......................................................... 8-6 Maintenance Program............................................................................. 8-6 Structural and Pressure Tests ...................................................... 8-6 Leak Tests ................................................................................... 8-7 Subsystem Maintenance .............................................................. 8-7 Shielding ...................................................................................... 8-8 Thermal ........................................................................................ 8-8 References ............................................................................................. 8-9 Page 8-i

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Acceptance Tests The following reviews, inspections, and tests shall be performed on the TN-40 packaging prior to initial transport. Many of these tests will be performed at the fabricators facility prior to delivery of the cask to the utility for use. Tests will be performed in accordance with written procedures approved by Transnuclear, Inc. For the TN-40 casks that have been fabricated, loaded and used for storage under 10CFR72 requirements, use of acceptance tests performed during their fabrication is also acceptable.

Visual Inspection Visual inspections are performed at the fabricator's facility prior to initial use to ensure that the packaging conforms to the drawings and specifications. The visual inspections include:

cleanliness inspections,

visual weld inspections as required by ASME Code [1],

inspection of sealing surface finish, and

dimensional inspections for conformance with the drawings included in Chapter 1 and referenced in the Certificate of Compliance.

The visual inspection includes verifying that all specified coatings are applied and the packaging is clean and free of cracks, pinholes, uncontrolled voids or other defects that could significantly reduce its effectiveness. To the maximum extent practical, weld inspection is performed in accordance with the applicable ASME code sections [1].

Dimensions and tolerances shown on the drawings provided in Chapter 1 are confirmed by measurements. The sealing surfaces on the flange, lid and covers are inspected to ensure that there are no gouges, cracks or scratches that could result in an unacceptable leakage.

Prior to shipping, the packaging will be inspected to ensure that it is in good physical condition. This inspection shall include verification that all accessible cask surfaces are free of grease, oil or other contaminants, and that all cask components are in an acceptable condition for use.

Page 8-1

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Structural and Pressure Tests The structural analyses performed on the packaging are presented in Chapter 2. To ensure that the packaging can perform its design function, the structural materials are chemically and physically tested to confirm that the required properties are met. To the maximum extent practical, welding is performed using qualified processes and qualified personnel, according to the ASME Boiler and the Pressure Vessel Code [1]. Base materials and welds are examined in accordance with the ASME Boiler and Pressure Vessel code requirements. NDE requirements for welds are specified on the drawings provided in Chapter 1. All NDE is performed in accordance with written and approved procedures. The inspection personnel are qualified in accordance with SNT-TC-1A [2].

The containment welds are designed, fabricated, tested and inspected in accordance with ASME B&PV Code Subsection NB. Alternatives to the code taken regarding the containment vessel are described in Chapter 2, Section 2.11. The basket is designed, fabricated, and inspected in accordance with the ASME B&PV Code Subsection NB Alternatives to the code taken regarding the basket are described in Section 2.11.

Welds of the noncontainment structure are inspected per the NDE acceptance criteria of ASME B&PV Code, Subsection NF.

The TN-40 fuel basket is designed, fabricated, and inspected in accordance with the ASME B&PV Code Subsection NB. Fusion weld tests as required are shown in drawing 10421-71-9.

The impact limiter attachment bolt material is tested to show the Charpy fracture toughness is at least 20 ft-lb at -20°F. The tie rod material is tested to show the Charpy impact test energy is at least 35 ft-lb at -20°F.

Pressure Tests Prior to initial use a pressure test is performed on the cask assembly at a pressure of 25 psig. This is slightly higher than 1.5 times the maximum normal operating pressure of 15.7 psig. The test pressure is held for a minimum of 10 minutes. The test is performed in accordance with ASME B&PV Code,Section III, Subsection NB, Paragraph NB-6200 or NB-6300. All visible joints/surfaces are examined for possible leakage after application of the pressure. Temporary gaskets and seals may be used in place of the metallic seals during the test.

In addition, a bubble leak test is performed at a pressure of 3-5 psig on the neutron shield enclosure (outer shell, outer shell top and bottom rings). The purpose of this test is to identify any potential leak passages in the enclosure welds. The bubble leak test pressure is greater than the relief valve set pressure.

Page 8-2

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Load Tests The lifting trunnions are designed to exceed 10CFR71.45(a) lifting requirements. A load test of 1.5 times the design lift load is applied to the trunnions for a period of ten minutes, to ensure that the trunnions can perform satisfactorily. This has been approved previously in the NRC safety evaluations for the dry cask storage at Prairie Island [4].

A force equal to 1.5 times the impact limiter weight will be applied to the lifting lugs of each limiter for a period of ten minutes. At the conclusion of the test, the impact limiter lifting lugs (including welds) will be:

a. Visually examined for defects and permanent deformations.

b. Examined by the liquid penetrant method for defects. Acceptance standards will be in accordance with Article NF-5350 of Section III of the ASME Boiler and Pressure Vessel Code.

Containment Leakage Tests Leakage tests are performed to verify containment of contents by the packaging prior to first use for transportation. These tests are usually performed using the helium mass spectrometer method. Alternative methods are acceptable, provided that the required sensitivity is achieved. The leakage test is performed in accordance with ANSI N14.5

[3]. The acceptance criterion is 1 x 10-4 ref cm3/s when tested to a sensitivity of 5x10-5 ref cm3/s or better. The personnel performing the leakage test are qualified in accordance with SNT-TC-1A [2].

Page 8-3

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Component Tests 8.1.4.1 Valves, Rupture Discs, and Fluid Transport Devices There are no valves in the packaging performing a safety related function. The TN-40 design incorporates quick-disconnect couplings for ease of draining and venting.

However, these couplings do not form part of the containment boundary. They are covered by bolted closures with metallic seals. There is no required acceptance test for these components.

8.1.4.2 Gaskets The lid and all the other containment penetrations are sealed using double metallic seals. The inner seal forms part of the containment boundary. Metallic seals are not temperature sensitive, and are therefore tested at room temperature. Metallic seals of the same type as those to be used for transport are installed for the fabrication leak test, described in Section 8.1.3. The tested seals are replaced before loading the packaging for storage or for transport. The replacement seals are then leak-tested at the time of storage closure and/or prior to transport as described in Chapter 7.

8.1.4.3 Impact Limiter Leakage Test The following test will be performed prior to initial use, after all the seal welds are completed on the impact limiter, to verify that the impact limiter wood will be protected from any moisture exchange with the environment.

Pressurize each impact limiter container to a pressure between 2 and 3 psig using helium. Test all the weld seams for leakage using a soap bubble test.

8.1.4.4 Functional Tests The following functional tests will be performed prior to first use of the cask. Generally these tests will be performed at the fabrication facility.

a. Installation and removal of the lid, penetration covers, and other fittings will be observed. Each component will be checked for difficulties in installation and removal. After removal, each component will be visually examined for indications of deformation, galling, improper functioning, etc. Any defects will be corrected prior to acceptance of the cask.

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

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TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Shielding Tests The analyses performed to ensure the shielding integrity are presented in Chapter 5.

The radial neutron shield is protected from damage or loss by the aluminum and steel enclosure. The neutron shield material is a proprietary, borated, reinforced polymer.

The primary function of the resin is to provide neutron shielding, which is performed primarily by the hydrogen content of the resin. The resin also provides some gamma shielding, which is a function of the overall resin density, and is not sensitive to composition.

The shielding performance of the resin can be verified adequately by chemical analysis and verification of density. Uniformity is assured by installation process control.

The following are acceptance values for density and chemical composition for the resin.

The values used in the shielding calculations of Chapter 5 are included for comparison.

Chapter 5 values Acceptance Testing Values Element nominal wt % Element wt % acceptance range (%)

H 5.05 H 5.05 -10 / +20 B 1.05 B 1.05 20 The minimum resin density in acceptance testing is 1.547 g/cm3. Resin composition or density test results which fall outside of this range will be evaluated to ensure that the shielding regulatory dose limits are not exceeded.

Density testing will be performed on every mixed batch of resin. Chemical analysis will be made on the first batch mixed with a given set of components, and thereafter whenever a new lot of one of the major components is introduced. Major components are aluminum oxide, zinc borate and the polyester resin, which combined make up 92%

of the resin by weight.

Qualification tests of the personnel and procedure used for mixing and pouring the polyester resin used for radial neutron shielding are performed. Qualification testing includes verification that the chemical composition and density are achieved, and the process is performed in such a manner as to prevent voids.

The TN-40 casks are currently fabricated/loaded under a site specific Part 72 license.

Cask surface dose rate measurements (both neutron and gamma dose rates) after loading of the cask are required by Part 72 Technical Specifications. Results of these measurements demonstrate the adequacy of the as-fabricated gamma and neutron shielding and can be used as test data. The shielding is not expected to lose its effectiveness under long term storage conditions based on prior experience with loaded storage casks. In addition, during storage of the spent fuel, the cask does not experience dynamic loads that could cause failure of the shielding. Thus, a periodic test during the storage usage is not performed.

Page 8-5

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 In addition, prior to transport of the package, gamma and neutron dose rate measurements are to be taken over the complete cask surface to demonstrate the continued performance of the shielding. This documents both the shielding design and durability of the shielding materials. These measurements also demonstrate the loaded cask meets DOT shipping requirements.

Periodic testing of the shielding is not required because the Part 71 transport license application (Chapter 1) limits the use of the TN-40 for a one-time use.

Neutron Absorber Tests Boral is the neutron absorber used for criticality control in the TN-40 basket. The neutron absorber plates may be monolithic, or they may consist of paired plates, one containing boron in the specified areal density, and the other composed of aluminum or aluminum alloy to make up the balance of the specified thickness and thermal conductance.

The TN-40 safety analyses do not rely upon the tensile strength of these materials. The radiation and temperature environment in the cask is not sufficiently severe to damage these materials.

The Boral neutron absorber material consists of a core of aluminum and boron carbide powders between two outer layers of aluminum. The criticality calculations take credit for 75% of the minimum specified B10 areal density of Boral.

Thermal Acceptance Tests The thermal evaluation presented in Chapter 3 is based on design configurations and thermal properties taken from industry recognized standards for the specified materials.

Based on the discussion in Section 3.4.7, thermal acceptance testing is not required for the TN-40 package prior to shipment.

Maintenance Program Structural and Pressure Tests The TN-40 cask will be used as a storage cask prior to use as a transport cask. If a loaded cask is taken from storage and prepared for transport, no load testing beyond the initial fabrication load test is required prior to shipment.

Page 8-6

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 Leak Tests After lid or vent/drain port cover removal, the affected metallic containment seals shall be replaced and the entire containment boundary tested prior to spent fuel shipment to show a leak rate less than 1x10-4 ref cm3/sec when tested to a sensitivity of 5x10-5 ref cm3/s or better per ANSI N14.5 [3]. These tests are usually performed using the helium mass spectrometer method. Alternative methods are acceptable, provided that the required sensitivity is achieved. Because the seals are used only once, the preshipment leak tests may be used to fulfill the ANSI N14.5 requirements for maintenance and periodic testing.

No leak tests are required prior to shipment of an empty TN-40 packaging.

Subsystem Maintenance 8.2.3.1 Fasteners The lid bolts and vent and drain cover bolts shall be inspected whenever they are removed for deformed or stripped threads. Damaged parts shall be evaluated for continued use and replaced as required. At a minimum, the lid bolts and vent and drain cover bolts shall be replaced at least once per fifty (50) shipments (round trip).

8.2.3.2 Impact Limiters A visual examination of the impact limiters before each shipment will be performed to ensure that the impact limiters have not been degraded between shipment. If there is no evidence of weld cracking or other damage which could result in water in-leakage, the wood will not be degraded. If there is visual damage, the impact limiter will be removed from service, repaired, if possible, and inspected for degradation of the wood.

Impact limiters will be leak tested once every five years to ensure that water has not entered the impact limiters. If the leak test indicates that the impact limiters have a leak, a humidity test will be performed to verify that there is no free water in the impact limiters. An impact limiter that has a leak will be removed from service and repaired.

8.2.3.3 Valves, Rupture Discs, and Gaskets on Containment Vessel If a port cover or the lid is removed, the seals are replaced if the cask is about to be loaded with or contains spent fuel. At the time the TN-40 is first converted from storage to transport use, it is most likely that the fuel that has been in storage will remain in the TN-40 for transport, and therefore, the lid will not be removed. In this case, the lid seal and seals of penetrations which have not been opened will not be replaced. The seals will be leak tested in accordance with Section 7.1.3 Step 7.

The metallic seals may be reused for transport of an empty TN-40 packaging.

Page 8-7

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 The leak test port (the overpressure port in the storage configuration) is closed by the overpressure transport cover with a single metallic seal. This flange and seal are not part of the containment boundary. The quick connect couplings in the vent and drain ports are not part of the containment boundary.

There are no valves or rupture discs on the TN-40 packaging containment.

Shielding There are no periodic tests or inspections required for the TN-40 shielding. Radiation surveys will be performed of the package exterior to ensure that the limits specified in 10 CFR 71.47 are met prior to shipment.

Thermal There are no periodic tests or inspections required for the TN-40 heat transfer components.

Page 8-8

TN-40 Transportation Packaging Safety Analysis Report Rev. 17, 11/23 References

1. ASME Boiler and Pressure Vessel Code,Section III, 1989.

2. SNT-TC-1A, American Society for Nondestructive Testing, Personnel Qualification and Certification in Nondestructive Testing, 1987.

3. ANSI N14.5-1987, Leakage Tests on Packages for Shipment of Radioactive Materials.

4. Letter from Beth A. Wetzel (NRC) to Roger O. Anderson (NSPC), Safety Evaluation and Safety Assessment Related to Dry Cask Storage at Prairie Island, June 12, 1995.

Page 8-9

Enclosure 3 to E-62614 AFFIDAVIT PURSUANT TO 10 CFR 2.390 State of Maryland:

County of HOWARD:

I, Prakash Narayanan, depose and say that I am Chief Technical Officer of TN Americas LLC, duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR 2.390 of the Commissions regulations for withholding this information.

The information for which proprietary treatment is sought is listed below:

Enclosure 1 - Portions of the TN-40 System SAR (Proprietary Version)

This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged, or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commissions regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

1) The information sought to be withheld from public disclosure involves portions of the SAR analyses and SAR drawings related to the design of the TN-40 System, which is owned and has been held in confidence by TN Americas LLC.

2) The information is of a type customarily held in confidence by TN Americas LLC and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.

3) Public disclosure of the information is likely to cause substantial harm to the competitive position of TN Americas LLC because the information consists of descriptions of the design and analysis of a radioactive material transportation system, the application of which provide a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their products position or impair the position of TN America LLCs product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

I declare that the statements set forth in this affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 12 / 05 / 2023 Prakash Narayanan Chief Technical Officer, TN Americas LLC Page 1 of 1

v t e Letter Sequence Request
EPID:L-2021-LLA-0086, Acceptance Review of the Model No. TN-40 Package (Open)
Initiation Request, Request Acceptance... Administration Questions 9 February 2023 9 February 2023 Answers Results Approval Other: ML22164A323, ML23347A137, ML23347A139
MONTHYEAR2023-12-18 [Table View]

ML23341A013

Text

Subject:

References:

Enclosures:

SUMMARY

SUMMARY

Reference:

Reference:

6.0 CRITICALITY EVALUATION

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