Text

Robatel Technologies, LLC, RT-100 Type B Cask Safety Analysis Report, Revision 10
	ML23181A131
Person / Time
Site:	07109365
Issue date:	06/20/2023
From:	; Robatel Technologies
To:	; Office of Nuclear Material Safety and Safeguards
Shared Package
ML23181A127	List: ML23181A128; ML23181A129; ML23181A130; ML23181A131; ML23181A134;
References
	Download: ML23181A131 (1)
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(Non-Proprietary Version)

RT-100 Type B Cask Safety Analysis Report Docket Number 71-9365 Revision 10 June 20, 2023 This Part 71 Application for Approval of RT-100 Type B Cask Package for Radioactive Material represents Robatel Technologies, LLC approach to its business as applied to the specifications of this submittal. This Application requests that the Nuclear Regulatory Commission respects the proprietary information and withholds it from public disclosure subject to the provisions of 10 CFR 2.390. All detailed drawings are considered proprietary information.

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 TABLE OF CONTENTS

1. GENERAL INFORMATION................................................................................................ 1-1 1.1 Introduction ...................................................................................................................... 1-1 1.2 Package Description ......................................................................................................... 1-3 1.2.1 Packaging ................................................................................................................................. 1-3 1.2.1.1 Overall Dimensions .......................................................................................................... 1-3 1.2.1.2 Weight ............................................................................................................................... 1-4 1.2.1.3 Containment Features ....................................................................................................... 1-4 1.2.1.4 Neutron and Gamma Shielding Features .......................................................................... 1-4 1.2.1.5 Shielding Features for Personnel Barriers ........................................................................ 1-4 1.2.1.6 Criticality Control Features............................................................................................... 1-4 1.2.1.7 Structural Features - Lifting and Tie-Down Devices........................................................ 1-4 1.2.1.8 Structural Features - Impact Limiters ............................................................................... 1-5 1.2.1.9 Structural Features - Internal Supporting or Positioning Features ................................... 1-5 1.2.1.10 Structural Features - Outer Shell or Outer Packaging .................................................... 1-5 1.2.1.11 Structural Features - Packaging Closure Device ............................................................ 1-6 1.2.1.12 Structural Features - Heat Transfer Features .................................................................. 1-6 1.2.1.13 Structural Features - Packaging Markings ..................................................................... 1-6 1.2.1.14 Additional Information ................................................................................................... 1-6 1.2.2 Contents .................................................................................................................................... 1-6 1.2.2.1 Identification and Maximum Quantity of Radioactive Material ....................................... 1-7 1.2.2.2 Identification and Maximum Quantity of Fissile Material ............................................... 1-7 1.2.2.3 Physical and Chemical Form - Density, Moisture Content and Moderators .................... 1-7 1.2.2.3.1 Ion-Exchange Resins ................................................................................................. 1-7 1.2.2.3.2 Filters ......................................................................................................................... 1-8 1.2.2.3.3 Activated Hardware ................................................................................................... 1-8 1.2.2.3.4 Secondary Containers ................................................................................................ 1-8 1.2.2.4 Location and Configuration .............................................................................................. 1-8 1.2.2.5 Use of Non-Fissile Materials as Neutron Absorbers/Moderators ..................................... 1-9 1.2.2.6 Chemical/Galvanic/Gas Generation .................................................................................. 1-9 1.2.2.7 Maximum Weight of Contents and Payload ................................................................... 1-10 1.2.2.8 Maximum Decay Heat .................................................................................................... 1-10 1.2.2.9 Loading Restrictions ....................................................................................................... 1-10 Robatel Technologies, LLC Page TOC-1

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.2.2.10 Contents for the Certificate of Compliance .................................................................. 1-10 1.2.3 Special Requirements for Plutonium ...................................................................................... 1-10 1.2.4 Operational Features............................................................................................................... 1-11 1.3 Engineering Drawings and Additional Information................................................... 1-11 1.3.1 Engineering Drawings ............................................................................................................ 1-11 1.3.2 Conformance to Approved Design ......................................................................................... 1-11 1.3.3 Referenced Pages ................................................................................................................... 1-11 1.3.4 Special Fabrication Procedures .............................................................................................. 1-11 1.3.5 Package Category ................................................................................................................... 1-11 1.3.6 Supplemental Information ...................................................................................................... 1-11 1.4 Appendix ......................................................................................................................... 1-12 1.5 References ....................................................................................................................... 1-27

2. STRUCTURAL EVALUATION .......................................................................................... 2-1 2.1 Description of Structural Design .................................................................................... 2-1 2.1.1 Discussion ................................................................................................................................ 2-3 2.1.1.1 Containment Boundary ..................................................................................................... 2-4 2.1.2 Design Criteria ......................................................................................................................... 2-4 2.1.2.1 Cask Body Criteria (except Bolts and O-Rings) ............................................................... 2-5 2.1.2.2 Bolts .................................................................................................................................. 2-5 2.1.2.3 Lead .................................................................................................................................. 2-6 2.1.2.4 Foam ................................................................................................................................. 2-6 2.1.3 Weights and Centers of Gravity ............................................................................................... 2-6 2.1.4 Identification of Codes and Standards for Package Design ..................................................... 2-7 2.2 Materials ........................................................................................................................... 2-8 2.2.1 Material Properties and Specifications ..................................................................................... 2-8 2.2.2 Chemical, Galvanic, or Other Reactions ................................................................................ 2-10 2.2.2.1 Component Material Categories ..................................................................................... 2-10 2.2.2.1.1 Stainless/Nickel Alloy Steels................................................................................... 2-11 2.2.2.1.2 Nonferrous Metals ................................................................................................... 2-11 2.2.2.1.3 Shielding Materials .................................................................................................. 2-11 2.2.2.1.4 Criticality Control Material ..................................................................................... 2-12 2.2.2.1.5 Energy Absorbing Material ..................................................................................... 2-12 2.2.2.1.6 Cellular Foam and Insulation .................................................................................. 2-12 Robatel Technologies, LLC Page TOC-2

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.2.2.1.7 Lubricant and Grease ............................................................................................... 2-12 2.2.2.1.8 O-Rings.................................................................................................................... 2-12 2.2.2.1.9 Secondary Containers and Shoring.......................................................................... 2-12 2.2.2.1.10 Filters ..................................................................................................................... 2-12 2.2.2.2 General Effects of Identified Reactions .......................................................................... 2-12 2.2.2.3 Adequacy of the Cask Operating Procedures.................................................................. 2-13 2.2.2.4 Effects of Reaction Products ........................................................................................... 2-13 2.2.3 Effects of Radiation on Materials ........................................................................................... 2-13 2.3 Fabrication and Examination........................................................................................ 2-13 2.3.1 Fabrication .............................................................................................................................. 2-13 2.3.2 Examination ........................................................................................................................... 2-13 2.4 General Requirements for All Packages ...................................................................... 2-14 2.4.1 Minimum Package Size .......................................................................................................... 2-14 2.4.2 Tamper-Indicating Feature ..................................................................................................... 2-14 2.4.3 Positive Closure...................................................................................................................... 2-14 2.5 Lifting and Tie-Down Standards for All Packages ..................................................... 2-14 2.5.1 Lifting Devices ....................................................................................................................... 2-14 2.5.1.1 Lifting Design Criteria .................................................................................................... 2-15 2.5.1.2 Lifting Device Descriptions ............................................................................................ 2-15 2.5.1.3 Lifting Device Evaluations ............................................................................................. 2-15 2.5.1.3.1 Cask Body Lifting Evaluation ................................................................................. 2-15 2.5.1.3.1.1 Lifting Pocket Design Features ........................................................................ 2-16 2.5.1.3.1.2 Lifting Pocket Tear-out Stresses ...................................................................... 2-17 2.5.1.3.1.3 Lifting Pocket Bearing Stresses ....................................................................... 2-18 2.5.1.3.1.4 Lifting Pocket Weld Stresses ........................................................................... 2-18 2.5.1.3.1.5 Lifting Pocket Average Pure Shear .................................................................. 2-20 2.5.1.3.1.6 Summary of Results ......................................................................................... 2-20 2.5.1.3.2 Primary Lid Lifting Evaluation ............................................................................... 2-21 2.5.1.3.2.1 Primary Lid Lifting Ring Working Loads ........................................................ 2-21 2.5.1.3.2.2 Primary Lid Thread Engagement ..................................................................... 2-22 2.5.1.3.3 Secondary Lid Lifting Evaluation ........................................................................... 2-23 2.5.1.3.3.1 Lifting Ring Working Load.............................................................................. 2-23 2.5.1.3.3.2 Secondary Lid Thread Engagement ................................................................. 2-23 Robatel Technologies, LLC Page TOC-3

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.5.1.3.4 Upper Impact Limiter Lifting Evaluation ................................................................ 2-24 2.5.1.3.4.1 Lifting Ring Working Load.............................................................................. 2-24 2.5.1.3.4.2 Impact Limiter Thread Engagement................................................................. 2-25 2.5.1.3.5 Lower Impact Limiter Lifting Evaluation ............................................................... 2-26 2.5.1.3.5.1 Attachment Bolt Stresses.................................................................................. 2-26 2.5.1.3.5.2 Lower Impact Limiter Thread Engagement ..................................................... 2-27 2.5.2 Tie-down Devices................................................................................................................... 2-28 2.5.2.1 Tie-down Load Calculation ............................................................................................ 2-28 2.5.2.2 Tie-down Force Calculation............................................................................................ 2-29 2.5.2.3 Tie-Down Arm Evaluation.............................................................................................. 2-32 2.5.2.4 Tie-down Arm & Plate Weld Evaluation ........................................................................ 2-33 2.5.2.4.1 Tie Down Arm-to-Plate Weld Stress........................................................................ 2-34 2.5.2.4.2 Tie Down Plate-to-Outer Shell Weld Stress ............................................................ 2-34 2.5.2.5 Tie-Down Evaluation Summary ..................................................................................... 2-35 2.6 Normal Conditions of Transport .................................................................................. 2-35 2.6.1 Heat ........................................................................................................................................ 2-36 2.6.1.1 Summary of Pressures and Temperatures ....................................................................... 2-36 2.6.1.2 Differential Thermal Expansion...................................................................................... 2-37 2.6.1.3 Stress Calculations .......................................................................................................... 2-37 2.6.1.4 Comparison with Allowable Stresses.............................................................................. 2-37 2.6.2 Cold ........................................................................................................................................ 2-37 2.6.3 Reduced External Pressure ..................................................................................................... 2-38 2.6.4 Increased External Pressure.................................................................................................... 2-38 2.6.5 Vibration................................................................................................................................. 2-38 2.6.5.1 Vibration Evaluation of the RT-100 Cask Primary Lid Bolts......................................... 2-38 2.6.5.2 Vibration Evaluation of the RT-100 Cask Secondary Lid Bolts ..................................... 2-40 2.6.6 Water Spray ............................................................................................................................ 2-41 2.6.7 Free Drop................................................................................................................................ 2-41 2.6.7.1 Methodology ................................................................................................................... 2-41 2.6.7.2 Finite Element Analysis .................................................................................................. 2-41 2.6.7.2.1 Model Description ................................................................................................... 2-42 2.6.7.2.2 Boundary Conditions ............................................................................................... 2-47 2.6.7.3 Side Drop ........................................................................................................................ 2-52 2.6.7.4 End Drop ......................................................................................................................... 2-60 Robatel Technologies, LLC Page TOC-4

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.6.8 Corner Drop............................................................................................................................ 2-68 2.6.9 Compression ........................................................................................................................... 2-68 2.6.10 Penetration ............................................................................................................................ 2-68 2.7 Hypothetical Accident Conditions ................................................................................ 2-68 2.7.1 Free Drop................................................................................................................................ 2-68 2.7.1.1 End Drop ......................................................................................................................... 2-70 2.7.1.1.1 End Drop Evaluation ............................................................................................... 2-70 2.7.1.1.2 Lead Slump Evaluation ........................................................................................... 2-70 2.7.1.1.2.1 Elastic Deformation.......................................................................................... 2-70 2.7.1.1.2.2 Plastic Deformation with Maximum Gap......................................................... 2-70 2.7.1.2 Side Drop ........................................................................................................................ 2-81 2.7.1.3 Corner Drop .................................................................................................................... 2-89 2.7.1.4 Oblique Drops ................................................................................................................. 2-91 2.7.1.5 Summary of Results ........................................................................................................ 2-93 2.7.2 Crush ...................................................................................................................................... 2-93 2.7.3 Puncture .................................................................................................................................. 2-93 2.7.3.1 Lid Puncture .................................................................................................................... 2-93 2.7.3.1.1 Lid Puncture Boundary Conditions ......................................................................... 2-93 2.7.3.1.2 Lid Puncture Results................................................................................................ 2-93 2.7.3.2 Cask Side Puncture ......................................................................................................... 2-96 2.7.3.2.1 Cask Side Puncture Minimum Wall Thickness ....................................................... 2-96 2.7.3.2.2 Cask Sidewall Bending Stresses .............................................................................. 2-96 2.7.3.3 Lead Deformation during Side Puncture ........................................................................ 2-97 2.7.3.3.1 Outer Shell Stiffness ................................................................................................ 2-97 2.7.3.3.2 Lead Stiffness .......................................................................................................... 2-98 2.7.3.3.3 Inner Shell Stiffness ................................................................................................ 2-98 2.7.3.3.4 Lead Deformation due to Puncture Load ................................................................ 2-98 2.7.4 Thermal ................................................................................................................................ 2-101 2.7.4.1 Summary of Pressures and Temperatures ..................................................................... 2-101 2.7.4.2 Differential Thermal Expansion.................................................................................... 2-101 2.7.4.3 Stress Calculations ........................................................................................................ 2-101 2.7.4.3.1 Bolt stresses during fire accident ........................................................................... 2-102 2.7.4.3.2 Pressure stress during fire accident........................................................................ 2-102 2.7.4.4 Comparison with Allowable Stresses............................................................................ 2-102 Robatel Technologies, LLC Page TOC-5

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.5 Immersion - Fissile Material ................................................................................................ 2-104 2.7.6 Immersion - All Package ..................................................................................................... 2-104 2.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2)........... 2-104 2.7.8 Summary of Damage ............................................................................................................ 2-104 2.8 Accident Conditions for Air Transport of Plutonium ............................................... 2-104 2.9 Accident Conditions for Fissile Material Packages for Air Transport..................... 2-104 2.10 Special Form ............................................................................................................... 2-104 2.11 Fuel Rods .................................................................................................................... 2-104 2.12 Appendix - Impact Limiter Analysis ....................................................................... 2-105 2.12.1 Assumptions ....................................................................................................................... 2-105 2.12.2 Analysis Inputs ................................................................................................................... 2-105 2.12.2.1 Cask Assembly............................................................................................................ 2-105 2.12.2.2 Foam Material Properties ............................................................................................ 2-107 2.12.2.2.1 Density................................................................................................................. 2-107 2.12.2.2.2 Crush Strength ..................................................................................................... 2-107 2.12.2.3 Temperatures............................................................................................................... 2-107 2.12.3 Methodology ...................................................................................................................... 2-108 2.12.3.1 Numerical Integration ................................................................................................. 2-109 2.12.3.2 Crush Strength ............................................................................................................ 2-110 2.12.3.3 Crush Force ................................................................................................................. 2-113 2.12.3.3.1 End-Drop Case .................................................................................................... 2-114 2.12.3.3.2 Side-Drop Case.................................................................................................... 2-116 2.12.3.3.3 Corner-Drop Case ................................................................................................ 2-119 2.12.4 Calculations ........................................................................................................................ 2-123 2.12.4.1 RT-100 Cask Drop Analysis ....................................................................................... 2-123 2.12.4.1.1 Calculation for Drop Height of 9.0 m.................................................................. 2-123 2.12.4.1.1.1 HAC End-Drop Case .................................................................................... 2-124 2.12.4.1.1.2 HAC Side-Drop Case ................................................................................... 2-127 2.12.4.1.1.3 HAC Corner-Drop Case ............................................................................... 2-130 2.12.4.1.2 Calculation for Drop Height of 0.3 m.................................................................. 2-134 2.12.4.1.2.1 NCT End-Drop Case .................................................................................... 2-134 2.12.4.1.2.2 NCT Side-Drop Case.................................................................................... 2-137 2.12.4.1.2.3 NCT Corner-Drop Case................................................................................ 2-140 Robatel Technologies, LLC Page TOC-6

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.12.4.2 Pin Puncture ................................................................................................................ 2-143 2.12.4.3 Confirmatory Testing - 3/10 Scale Cask ..................................................................... 2-144 2.12.4.3.1 End-Drop ............................................................................................................. 2-146 2.12.4.3.2 Corner-Drop ........................................................................................................ 2-148 2.12.4.3.3 Side-Drop ............................................................................................................ 2-150 2.12.5 Discussion of Test Results ................................................................................................. 2-152 2.12.6 Summary of Accelerations ................................................................................................. 2-152 2.13 Appendix - Closure Bolt Evaluation ........................................................................ 2-161 2.13.1 Methodology ...................................................................................................................... 2-161 2.13.2 Loads .................................................................................................................................. 2-161 2.13.2.1 Internal Pressure Loads ............................................................................................... 2-161 2.13.2.1.1 Internal Pressure Loads for Primary Lid Closure Bolts....................................... 2-161 2.13.2.1.2 HAC (Fire) Internal Pressure Load for Primary Lid Closure Bolts ..................... 2-163 2.13.2.1.3 Internal Pressure Load for Secondary Lid Closure Bolts .................................... 2-164 2.13.2.1.4 HAC (Fire) Internal Pressure Load for Secondary Lid Closure Bolts ................. 2-165 2.13.2.2 Temperature Loads ..................................................................................................... 2-166 2.13.2.2.1 NCT Temperature Loads for Primary Lid Closure Bolts .................................... 2-166 2.13.2.2.2 HAC (Fire) Temperature Loads for Primary Lid Closure Bolts.......................... 2-167 2.13.2.2.3 NCT Temperature Loads for Secondary Lid Closure Bolts ................................ 2-168 2.13.2.2.4 HAC (Fire) Temperature Loads for Secondary Lid Closure Bolts...................... 2-168 2.13.2.3 Bolt Preloads ............................................................................................................... 2-168 2.13.2.3.1 Bolt Preload for Primary Lid Closure Bolts ........................................................ 2-168 2.13.2.3.2 Bolt Preload for Secondary Lid Closure Bolts .................................................... 2-169 2.13.2.4 Impact Loads............................................................................................................... 2-170 2.13.2.4.1 Dynamic Load Factors ........................................................................................ 2-170 2.13.2.4.2 NCT End Drop Loads .......................................................................................... 2-173 2.13.2.4.2.1 Primary Lid Bolts ......................................................................................... 2-173 2.13.2.4.2.2 Secondary Lid Bolts ..................................................................................... 2-174 2.13.2.4.3 HAC End Drop Loads ......................................................................................... 2-176 2.13.2.4.3.1 Primary Lid Bolts ......................................................................................... 2-176 2.13.2.4.3.2 Secondary Lid Bolts ..................................................................................... 2-177 2.13.2.4.4 Corner Drop Evaluations ..................................................................................... 2-179 2.13.2.4.5 Side Drop Evaluations ......................................................................................... 2-179 2.13.2.5 Puncture Loads............................................................................................................ 2-179 Robatel Technologies, LLC Page TOC-7

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.2.5.1 End Puncture ....................................................................................................... 2-180 2.13.2.5.1.1 Primary Lid Bolts ......................................................................................... 2-180 2.13.2.5.1.2 Secondary Lid Bolts ..................................................................................... 2-183 2.13.2.5.2 Side Puncture ....................................................................................................... 2-186 2.13.2.6 External Pressure ........................................................................................................ 2-186 2.13.2.6.1 Primary Lid Bolts ................................................................................................ 2-187 2.13.2.6.2 Secondary Lid Bolts ............................................................................................ 2-188 2.13.2.7 Gasket Seating Load ................................................................................................... 2-189 2.13.2.8 Vibration-induced Loads ............................................................................................ 2-189 2.13.2.8.1 NCT Vibration-induced Loads for Primary Lid Closure Bolts ........................... 2-189 2.13.2.8.2 NCT Vibration-induced Loads for Secondary Lid Closure Bolts ....................... 2-190 2.13.3 Load Combinations ............................................................................................................ 2-191 2.13.3.1 NCT Load Combination for Primary Lid Bolt............................................................ 2-193 2.13.3.2 NCT Load Combination for Secondary Lid Bolt........................................................ 2-197 2.13.3.3 HAC Load Combination for Primary Lid Bolt ........................................................... 2-202 2.13.3.4 HAC Load Combination for Secondary Lid Bolt ....................................................... 2-204 2.13.4 Fatigue Analysis ................................................................................................................. 2-206 2.13.4.1 Fatigue Analysis for Primary Lid Bolts ...................................................................... 2-207 2.13.4.1.1 Normal Operation Cycles .................................................................................... 2-207 2.13.4.1.2 Vibration Cycles .................................................................................................. 2-208 2.13.4.2 Fatigue Analysis for Secondary Lid Bolts .................................................................. 2-209 2.13.4.2.1 Normal Operation Cycles .................................................................................... 2-209 2.13.4.2.2 Vibration Cycles .................................................................................................. 2-209 2.13.5 Seal Integrity ...................................................................................................................... 2-210 2.13.5.1 Primary Lid Seals ........................................................................................................ 2-211 2.13.5.2 Secondary Lid Seals .................................................................................................... 2-211 2.13.6 Vent Port Cover Plate O-Ring and Bolt Evaluation ........................................................... 2-212 2.13.6.1 Vent Port Cover Plate O-Ring Evaluation .................................................................. 2-212 2.13.6.1.1 O-ring Sealing Force ........................................................................................... 2-212 2.13.6.1.2 Vent Port Cover Plate Preload ............................................................................. 2-213 2.13.6.1.3 Factor of Safety to Maintain a Tight Seal ............................................................ 2-213 2.13.6.2 Bolt Evaluation ........................................................................................................... 2-213 2.13.6.2.1 Thread Engagement ............................................................................................. 2-214 2.13.6.2.2 Thread Shear Evaluation ..................................................................................... 2-214 Robatel Technologies, LLC Page TOC-8

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.6.2.3 Load to Break Bolt .............................................................................................. 2-215 2.14 Appendix - Fabrication Stress Evaluation .............................................................. 2-215 2.14.1 Lead Pour ........................................................................................................................... 2-216 2.14.1.1 Cask Shell Geometry .................................................................................................. 2-216 2.14.1.2 Stresses Resulting from Lead Pour ............................................................................. 2-216 2.14.2 Cool-down .......................................................................................................................... 2-217 2.14.2.1 Hoop (Circumferential) Stresses ................................................................................. 2-217 2.14.2.2 Axial Stress ................................................................................................................. 2-219 2.14.2.3 Effects of Temperature Differential during Cool-down.............................................. 2-220 2.14.3 Lead Creep ......................................................................................................................... 2-220 2.15 Appendix - Seal Region Stress Evaluation .............................................................. 2-220 2.15.1 Seal Region Post-Processing Methodology........................................................................ 2-220 2.15.2 Stress Concentration Factors .............................................................................................. 2-221 2.15.3 Seal Region Stress Results ................................................................................................. 2-221 2.15.4 Displacement Results ......................................................................................................... 2-221 2.16 Appendix - Design Fatigue Curves for High Strength Steel Bolting [Ref. 66] .... 2-233 2.17 References ................................................................................................................... 2-234

3. THERMAL EVALUATION ................................................................................................. 3-1 3.1 Description of Thermal Design........................................................................................ 3-3 3.1.1 Design Features ........................................................................................................................ 3-3 3.1.1.1 RT-100 Description .......................................................................................................... 3-3 3.1.1.2 RT-100 Dimensions .......................................................................................................... 3-4 3.1.2 Contents Decay Heat ............................................................................................................... 3-4 3.1.3 Summary Tables of Temperatures............................................................................................ 3-5 3.1.4 Summary Tables of Maximum Pressures ................................................................................. 3-7 3.2 Material Properties and Component Specifications ...................................................... 3-7 3.2.1 Material Properties ................................................................................................................... 3-8 3.2.2 Component Specifications ...................................................................................................... 3-11 3.2.3 Content Properties .................................................................................................................. 3-12 3.3 Thermal Evaluation under Normal Conditions of Transport .................................... 3-13 3.3.1 Heat and Cold ......................................................................................................................... 3-14 3.3.1.1 Load Cases ...................................................................................................................... 3-14 3.3.1.2 Analytical Model ............................................................................................................ 3-16 Robatel Technologies, LLC Page TOC-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.3.1.3 Analysis Results .............................................................................................................. 3-19 3.3.2 Maximum Normal Operating Pressure ................................................................................... 3-29 3.3.2.1 Calculation Method......................................................................................................... 3-29 3.3.2.2 Pressure Due to the Initially Sealed Air in the Cavity .................................................... 3-29 3.3.2.3 Pressure Due to the Water Vapor in the Cask ................................................................. 3-29 3.3.2.4 Pressure Due to Generation of Gas ................................................................................. 3-30 3.3.2.5 Total Pressure.................................................................................................................. 3-30 3.4 Thermal Evaluation under Hypothetical Accident Conditions .................................. 3-31 3.4.1 HAC Fire AnalysisPin Puncture Damage to Top Impact Limiter ...................................... 3-31 3.4.1.1 Initial ConditionsPin Puncture Damage to Top Impact Limiter ................................. 3-31 3.4.1.2 HAC Fire and Cool-down AnalysisPin Puncture Damage to Top Impact Limiter ........................................................................................................................... 3-31 3.4.1.3 HAC Fire Analysis ResultsPin Puncture Damage to Top Impact Limiter .................. 3-32 3.4.2 HAC Fire EvaluationPin Puncture Damage to the Side of the Cask Body ........................ 3-43 3.4.2.1 Initial ConditionPin Puncture Damage to the Side of the Cask Body ........................ 3-43 3.4.2.2 HAC Fire AnalysisPin Puncture Damage to the Side of the Cask Body .................... 3-44 3.4.2.3 HAC Fire and Cool-down AnalysisPin Puncture Damage to the Side of the Cask Body ...................................................................................................................... 3-44 3.4.3 Maximum Temperatures and Pressure ................................................................................... 3-57 3.4.3.1 Maximum Temperatures ................................................................................................. 3-57 3.4.3.2 Maximum Accident Condition Pressure ......................................................................... 3-57 3.4.3.2.1 Calculation Method ................................................................................................. 3-58 3.4.3.2.2 Pressure Due to the Initially Sealed Air in the Cavity ............................................. 3-58 3.4.3.2.3 Pressure Due to the Water Vapor in the Cask ......................................................... 3-58 3.4.3.2.4 Pressure Due to Generation of Gas.......................................................................... 3-58 3.4.3.2.5 Total Pressure .......................................................................................................... 3-59 3.4.3.2.6 Total Pressure Accounting for Combustion of Contents ......................................... 3-59 3.4.4 Maximum Thermal Stress ...................................................................................................... 3-60 3.4.5 Accident Conditions for Fissile Material Packages for Air Transport ................................... 3-60 3.5 Appendix ......................................................................................................................... 3-61 3.6 References ....................................................................................................................... 3-70

4. CONTAINMENT EVALUATION....................................................................................... 4-1 4.1 Description of Containment System ............................................................................... 4-1 Robatel Technologies, LLC Page TOC-10

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.1.1 Containment Vessel .................................................................................................................. 4-1 4.1.2 Containment Penetration .......................................................................................................... 4-3 4.1.3 Welds and Seals........................................................................................................................ 4-5 4.1.4 Closure ..................................................................................................................................... 4-5 4.1.5 Cavity Volume, Conditions, and Contents ............................................................................... 4-6 4.2 Allowable Leakage Rates at Test Conditions ................................................................ 4-7 4.3 Leakage Rate Test for Type B Packages ....................................................................... 4-8 4.3.1 Determination of Equivalent Reference Leakage Rate for Helium Gas................................... 4-9 4.3.2 Determination of Equivalent Reference Leakage Rate for Air .............................................. 4-15 4.4 Hydrogen Gas Generation ............................................................................................ 4-16 4.4.1 Determination of Bounding G Values .................................................................................... 4-17 4.4.1.1 G Values for Waste and Secondary Container Materials................................................ 4-17 4.4.1.2 Calculation of Effective G Values .................................................................................. 4-19 4.4.1.3 Operating Temperature G Value Adjustment ................................................................. 4-20 4.4.2 Hydrogen Gas Generation by Radiolysis ............................................................................... 4-22 4.4.3 Hydrogen Generation - Radiolysis in Waste, Water and Polyethylene Container ................ 4-24 4.4.4 Hydrogen Gas Generation - Simplified Model used to develop Loading Curve ................... 4-31 4.4.5 Hydrogen Gas Generation - Analytical Model used for Detailed Analysis ........................... 4-34 4.5 Appendix ......................................................................................................................... 4-37 4.6 References ....................................................................................................................... 4-48

5. SHIELDING EVALUATION ............................................................................................... 5-1 5.1 Description of Shielding Design ...................................................................................... 5-5 5.1.1 Design Features ........................................................................................................................ 5-5 5.1.2 Summary Table of Maximum Radiation Levels ...................................................................... 5-5 5.2 Source Specification ......................................................................................................... 5-6 5.2.1 Gamma Source ......................................................................................................................... 5-6 5.2.2 Neutron Source ......................................................................................................................... 5-7 5.3 Shielding Model ................................................................................................................ 5-8 5.3.1 Configuration of Source and Shielding .................................................................................... 5-8 5.3.1.1 Source Term ...................................................................................................................... 5-9 5.3.1.2 NCT Model ....................................................................................................................... 5-9 5.3.1.3 HAC Model....................................................................................................................... 5-9 5.3.2 Material Properties ................................................................................................................. 5-17 Robatel Technologies, LLC Page TOC-11

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4 Shielding Evaluation - Resins and Filters ................................................................ 5-19 5.4.1 Methods .................................................................................................................................. 5-19 5.4.1.1 MCNP6 Analysis ............................................................................................................ 5-19 5.4.1.2 Dose Rate Response Calculation .................................................................................... 5-19 5.4.1.3 Maximum allowed source strength density .................................................................... 5-23 5.4.1.4 Variance Reduction......................................................................................................... 5-24 5.4.2 Input and Output Data ............................................................................................................ 5-24 5.4.3 Flux-to-Dose Rate Conversion ............................................................................................... 5-24 5.4.4 External Radiation Levels ...................................................................................................... 5-26 5.4.4.1 MCNP6 Statistics Evaluation ......................................................................................... 5-27 5.4.4.1.1 Tally Statistics Diagnostics ..................................................................................... 5-27 5.4.4.1.2 Fractional Standard Deviation of Individual Tally Segments ................................. 5-27 5.4.4.2 Media Composition and Density..................................................................................... 5-29 5.4.4.2.1 Effect of Media Composition .................................................................................. 5-29 5.4.4.2.2 Effect of Media Density .......................................................................................... 5-30 5.4.4.3 Shielding Evaluation Uncertainty ................................................................................... 5-31 5.4.4.3.1 Calculation of Dose Rates ....................................................................................... 5-31 5.4.4.3.1.1 Radiation Source Generation............................................................................ 5-31 5.4.4.3.1.2 Cross Section Data ........................................................................................... 5-32 5.4.4.3.1.3 Radiation Transport Codes ............................................................................... 5-32 5.4.4.3.2 Attenuation from other material (i.e. secondary containers) not included in shielding analysis .................................................................................................. 5-32 5.4.4.3.3 Nominal Media Bulk Density .................................................................................. 5-33 5.4.4.4 Loading Table ................................................................................................................. 5-33 5.4.4.5 Dose Rates for Maximum Radionuclide Loading ........................................................... 5-34 5.5 Shielding Evaluation - Mass Restricted Resins and Filters ....................................... 5-49 5.5.1 Radiation Level Summary ...................................................................................................... 5-49 5.5.2 Source Specification ............................................................................................................... 5-50 5.5.3 Shielding Model ..................................................................................................................... 5-50 5.5.4 Geometry and Materials ......................................................................................................... 5-50 5.5.5 Tally Structure ........................................................................................................................ 5-51 5.5.6 Specific Activity Limits ......................................................................................................... 5-55 5.5.7 Dose Rate Compliance ........................................................................................................... 5-60 Robatel Technologies, LLC Page TOC-12

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.6 Shielding Evaluation - Activated Hardware ............................................................... 5-61 5.6.1 Radiation Level Summary ...................................................................................................... 5-61 5.6.2 Source Specification ............................................................................................................... 5-62 5.6.3 Shielding Model ..................................................................................................................... 5-62 5.6.4 Geometry and Materials ......................................................................................................... 5-62 5.6.5 Tally Structure ........................................................................................................................ 5-66 5.6.6 Specific Activity Limits ......................................................................................................... 5-66 5.6.7 Dose Rate Compliance ........................................................................................................... 5-77 5.7 Appendix ......................................................................................................................... 5-79 5.7.1 List of Gamma Radionuclides with Greater than 1 Day Half Life ......................................... 5-79 5.7.2 Gamma Radionuclide Responses ........................................................................................... 5-81 5.7.3 Radionuclide Maximum Ci/g Loading Limits........................................................................ 5-97 5.7.4 Process Description for Calculating Maximum Allowed Source Strength Density ............. 5-105 5.8 References ..................................................................................................................... 5-106

6. CRITICALITY EVALUATION .......................................................................................... 6-1

7. PACKAGE OPERATIONS .................................................................................................. 7-1 7.1 Package Loading .............................................................................................................. 7-3 7.1.1 Preparation for Loading............................................................................................................ 7-3 7.1.1.1 Upper Impact Limiter Removal ........................................................................................ 7-5 7.1.1.2 Optional Loading Steps..................................................................................................... 7-5 7.1.1.3 Removal of Quick-Disconnect Valve Cover Plate ........................................................... 7-5 7.1.1.4 Removal of the Primary Lid.............................................................................................. 7-6 7.1.1.5 Removal of the Secondary Lid.......................................................................................... 7-6 7.1.2 Loading of the RT-100 ............................................................................................................. 7-6 7.1.2.1 Content Loading................................................................................................................ 7-7 7.1.2.2 Primary Lid Replacement ................................................................................................. 7-8 7.1.2.3 Secondary Lid Replacement ............................................................................................. 7-8 7.1.2.4 Quick-Disconnect Valve Cover Plate Replacement ......................................................... 7-8 7.1.3 Preparation for Transport ......................................................................................................... 7-9 7.1.3.1 Replacement of Upper Impact Limiter ........................................................................... 7-10 7.1.3.2 Verification for Transport ............................................................................................... 7-11 7.2 Package Unloading ........................................................................................................ 7-11 7.2.1 Receipt of Package from Carrier ............................................................................................ 7-11 Robatel Technologies, LLC Page TOC-13

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.2.2 Removal of Contents .............................................................................................................. 7-12 7.3 Preparation of Empty Package for Transport ............................................................ 7-13 7.4 Other Operations ........................................................................................................... 7-13 7.4.1 Lower Impact Limiter Removal ............................................................................................. 7-13 7.4.2 Package Removal from Trailer............................................................................................... 7-13 7.4.3 Replacement of Lower Impact Limiter .................................................................................. 7-16 7.4.4 Reloading the Package onto the Trailer.................................................................................. 7-16 7.4.5 Tightening of Components ..................................................................................................... 7-18 7.4.5.1 Tightening Torques ......................................................................................................... 7-18 7.4.5.2 Threaded Bolts - Tightening Methods and Equipment .................................................. 7-19 7.4.5.3 Replacement of the Impact Limiter Threaded Studs....................................................... 7-19 7.5 Hydrogen Buildup in RT-100 Transport Cask ........................................................... 7-20 7.5.1 Hydrogen Gas Generation - Simplified Model used to develop Loading Curve ................... 7-20 7.5.2 Hydrogen Gas Generation - Analytical Model used for Detailed Analysis ........................... 7-24 7.5.3 Hydrogen Gas Generation - Analytical Model Examples ..................................................... 7-27 7.6 Appendix ......................................................................................................................... 7-29 7.6.1 RT-100 Loading Table ........................................................................................................... 7-30 7.6.1.1 RT-100 Loading Table Description ................................................................................ 7-30 7.6.1.2 RT-100 Loading Table Procedure................................................................................... 7-35 7.6.1.3 Turkey Point Source Term Example Evaluation (Resin and Filter) ............................... 7-37 7.6.1.4 St. Lucie Loading Table (Resin and Filter)..................................................................... 7-39 7.6.1.5 Mixed Shipment Example (Resin/filter and Activated Hardware) ................................. 7-41 7.6.1.6 Additional Examples ....................................................................................................... 7-43 7.7 References ....................................................................................................................... 7-54

8. ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ......................................... 8-1 8.1 Acceptance Tests .............................................................................................................. 8-2 8.1.1 Visual Inspections and Measurements ..................................................................................... 8-3 8.1.2 Weld Examinations .................................................................................................................. 8-3 8.1.3 Structural and Pressure Tests ................................................................................................... 8-3 8.1.4 Leakage Tests ........................................................................................................................... 8-4 8.1.4.1 Cask Containment Boundary ............................................................................................ 8-5 8.1.4.1.1 Cask Body Leak Testing - Prior to Lead Pouring ..................................................... 8-5 8.1.4.1.2 Primary Lid Assembly Including Secondary Lid and Cover Plate - Prior to Robatel Technologies, LLC Page TOC-14

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Final Assembly ....................................................................................................... 8-6 8.1.4.2 Primary and Secondary Lid Containment O-Rings Helium Leak Testing........................ 8-7 8.1.4.3 Quick Disconnect Valve Helium Leak Testing ................................................................ 8-8 8.1.4.4 Quick Disconnect Valve Cover Plate Containment O-Rings Helium Leak Testing....... 8-10 8.1.5 Component and Material Tests............................................................................................... 8-12 8.1.5.1 Foam ............................................................................................................................... 8-12 8.1.5.2 O-Ring ............................................................................................................................ 8-12 8.1.5.3 Ceramic Paper ................................................................................................................. 8-13 8.1.5.4 Fusible Plugs ................................................................................................................... 8-14 8.1.5.5 Carbon Steel and Alloy Steel Fasteners .......................................................................... 8-14 8.1.5.6 Stainless Steel Fasteners ................................................................................................. 8-17 8.1.5.7 Thread Inserts.................................................................................................................. 8-19 8.1.5.8 Quick Disconnect Valve ................................................................................................. 8-20 8.1.6 Shielding Tests ....................................................................................................................... 8-20 8.1.7 Thermal Tests ......................................................................................................................... 8-21 8.1.8 Miscellaneous Tests ............................................................................................................... 8-21 8.2 Maintenance Program ................................................................................................... 8-21 8.2.1 Structural and Pressure Tests ................................................................................................. 8-21 8.2.2 Leakage Tests ......................................................................................................................... 8-22 8.2.2.1 Periodic and Maintenance Leak Test .............................................................................. 8-22 8.2.2.2 Pre-Shipment Leak Test - Gas Pressure Rise Option ..................................................... 8-22 8.2.2.3 Pre-Shipment Leak Test - Gas Pressure Drop Option .................................................... 8-25 8.2.3 Component and Material Tests............................................................................................... 8-27 8.2.3.1 Routine Component Inspection....................................................................................... 8-27 8.2.3.2 Annual Component Inspection ....................................................................................... 8-28 8.2.4 Thermal Tests ......................................................................................................................... 8-28 8.2.5 Miscellaneous Tests ............................................................................................................... 8-29 8.3 Appendix ......................................................................................................................... 8-30 8.3.1 Summary of Leak Test Requirements .................................................................................... 8-30 8.3.2 Minimum Lead Thickness and Gap Determination ............................................................... 8-31 8.3.2.1 Explanation of the Gap Between Lead and the External Shell ....................................... 8-31 8.3.2.2 Conclusion ...................................................................................................................... 8-36 8.4 References ....................................................................................................................... 8-37 Robatel Technologies, LLC Page TOC-15

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 List of Figures Figure 1-1 Information Flow for General Information ................................................................. 1-2 Figure 1.2.1-1 RT-100 Cask Package Artist Concept ........................................................................ 1-3 Figure 2-1 Information Flow for the Structural Review............................................................... 2-2 Figure 2.5.1-1 RT-100 Lifting Pocket Dimensions .......................................................................... 2-17 Figure 2.5.1-2 Weld Geometry......................................................................................................... 2-19 Figure 2.5.2-1 RT-100 Tie-Down Arm Geometry ............................................................................ 2-30 Figure 2.5.2-2 RT-100 Tie-Down Free Body Diagrams ................................................................... 2-31 Figure 2.6.7-1 RT-100 Solid Model ................................................................................................. 2-44 Figure 2.6.7-2 RT-100 Finite Element Model .................................................................................. 2-45 Figure 2.6.7-3 Gap Elements Used to Represent the Impact Limiters for Side and End Drop Configurations .......................................................................................................... 2-46 Figure 2.6.7-4 Bolt Pre-load Using ANSYS Pre-Tension Elements (PRETS179) ........................... 2-50 Figure 2.6.7-5 Pressure Distribution Used to Simulate the Contents ............................................... 2-51 Figure 2.6.7-6 RT-100 NCT Side Drop Stress Intensity Results ...................................................... 2-54 Figure 2.6.7-7 RT-100 Inner Shell NCT Side Drop Stress Intensity Results .................................... 2-55 Figure 2.6.7-8 RT-100 Outer Shell NCT Side Drop Stress Intensity Results .................................. 2-56 Figure 2.6.7-9 RT-100 Flange NCT Side Drop Stress Intensity Results .......................................... 2-57 Figure 2.6.7-10 RT-100 Outer Lid NCT Side Drop Stress Intensity Results ...................................... 2-58 Figure 2.6.7-11 RT-100 Inner Lid NCT Side Drop Stress Intensity Results ...................................... 2-59 Figure 2.6.7-12 RT-100 NCT Bottom Drop Stress Intensity Results ................................................. 2-62 Figure 2.6.7-13 RT-100 Inner Shell NCT End Drop Stress Intensity Results .................................... 2-63 Figure 2.6.7-14 RT-100 Outer Shell NCT End Drop Stress Intensity Results.................................... 2-64 Figure 2.6.7-15 RT-100 Flange NCT End Drop Stress Intensity Results ........................................... 2-65 Figure 2.6.7-16 RT-100 Outer Lid NCT End Drop Stress Intensity Results ...................................... 2-66 Figure 2.6.7-17 RT-100 Inner Lid NCT End Drop Stress Intensity Results ....................................... 2-67 Figure 2.7.1-1 RT-100 HAC End Drop Stress Intensity Results ...................................................... 2-74 Figure 2.7.1-2 RT-100 Inner Shell HAC End Drop Stress Intensity Results .................................... 2-75 Figure 2.7.1-3 RT-100 Outer Shell HAC End Drop Stress Intensity Results ................................... 2-76 Figure 2.7.1-4 RT-100 Flange HAC End Drop Stress Intensity Results........................................... 2-77 Figure 2.7.1-5 RT-100 Outer Lid HAC End Drop Stress Intensity Results ...................................... 2-78 Figure 2.7.1-6 RT-100 Inner Lid HAC End Drop Stress Intensity Results....................................... 2-79 Figure 2.7.1-7 RT-100 Lead Slump .................................................................................................. 2-80 Figure 2.7.1-8 RT-100 HAC Side Drop Stress Intensity Results ...................................................... 2-83 Robatel Technologies, LLC Page TOC-16

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-9 RT-100 Inner Shell HAC Side Drop Stress Intensity Results ................................... 2-84 Figure 2.7.1-10 RT-100 Outer Shell HAC Side Drop Stress Intensity Results................................... 2-85 Figure 2.7.1-11 RT-100 Flange HAC Side Drop Stress Intensity Results .......................................... 2-86 Figure 2.7.1-12 RT-100 Outer Lid HAC Side Drop Stress Intensity Results ..................................... 2-87 Figure 2.7.1-13 RT-100 Inner Lid HAC Side Drop Stress Intensity Results ...................................... 2-88 Figure 2.7.3-1 RT-100 ANSYS Puncture Model ............................................................................. 2-95 Figure 2.7.3-2 RT-100 Pin Puncture Stress Intensity Results ........................................................... 2-95 Figure 2.7.3-3 RT-100 Side Puncture Details ................................................................................ 2-100 Figure 2.12.2-1 Schematic of the RT100 impact limiters ................................................................ 2-106 Figure 2.12.3-1 Example Crush Strength-Strain Curve ................................................................... 2-113 Figure 2.12.3-2 Cask configuration at impact (End-Drop) .............................................................. 2-114 Figure 2.12.3-3 Cask configuration when velocity becomes zero (End-Drop) ................................ 2-115 Figure 2.12.3-4 Diagram for contact area calculation (End-Drop) .................................................. 2-115 Figure 2.12.3-5 Cask configuration at impact (Side-Drop) .............................................................. 2-117 Figure 2.12.3-6 Cask configuration when velocity becomes zero (Side-Drop) ............................... 2-117 Figure 2.12.3-7 Diagram of Contact Zones ...................................................................................... 2-118 Figure 2.12.3-8 Cask Configuration at Impact (Corner-Drop)......................................................... 2-119 Figure 2.12.3-9 Cask configuration when velocity becomes zero (Corner-Drop) ........................... 2-120 Figure 2.12.3-10 Typical Unit Cell Configuration ............................................................................. 2-121 Figure 2.12.3-11 Crush sequence of two foam cell ............................................................................ 2-122 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

Figure 2.13.5-1 Compression Set vs. Temperature........................................................................... 2-212 Figure 2.15.4-1 Stress Intensity Contour Plot of Primary Lid Following End Drop. ...................... 2-224 Figure 2.15.4-2 Stress Intensity Contour Plot of the Primary Seal Region ...................................... 2-225 Figure 2.15.4-3 Lid Seal Geometry .................................................................................................. 2-226 Figure 2.15.4-4 Primary Lid Sealing Surface Displacement during Side drop ................................ 2-227 Figure 2.15.4-5 Secondary Lid Sealing Surface Displacement during Side drop ............................ 2-228 Figure 2.15.4-6 Primary Lid Sealing Surface Displacement during End drop................................. 2-229 Figure 2.15.4-7 Secondary Lid Sealing Surface Displacement during End drop............................. 2-230 Figure 2.15.4-8 Primary Lid Sealing Surface Displacement during Puncture ................................. 2-231 Figure 2.15.4-9 Secondary Lid Sealing Surface Displacement during Puncture ............................. 2-232 Figure 3-1 Information Flow for the Thermal Review ................................................................. 3-2 Figure 3.3.1-1 RT-100 ANSYS Finite Element Model Volumes .................................................... 3-17 Figure 3.3.1-2 RT-100 ANSYS Normal Condition Finite Element Mesh...................................... 3-18 Figure 3.3.1-3 Temperature Contour Plot of PackageHot Case 1 ................................................ 3-19 Figure 3.3.1-4 Temperature Contour Plot of Cask BodyHot Case 1 ............................................ 3-20 Figure 3.3.1-5 Temperature Contour Plot of Inner Shell SurfaceHot Case 1............................... 3-21 Figure 3.3.1-6 Temperature Contour Plot of Lead ShieldingHot Case 1 ..................................... 3-22 Figure 3.3.1-7 Temperature Contour Plot of PackageHot Case 2 ................................................ 3-23 Robatel Technologies, LLC Page TOC-18

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-8 Temperature Contour Plot of Cask BodyHot Case 2 ............................................ 3-24 Figure 3.3.1-9 Temperature Contour Plot of PackageCold Case 1 .............................................. 3-25 Figure 3.3.1-10 Temperature Contour Plot of Cask BodyCold Case 1 .......................................... 3-26 Figure 3.3.1-11 Temperature Contour Plot of PackageCold Case 2 .............................................. 3-27 Figure 3.3.1-12 Temperature Contour Plot of Cask Body Cold Case 2 ............................................. 3-28 Figure 3.4.1-1 Temperature Contour Plot of Package Pre-Fire Fire ConditionHAC Pin Damage on Top Impact Limiter ............................................................................... 3-33 Figure 3.4.1-2 Temperature Contour Plot of Cask Body Pre-Fire ConditionHAC Pin Damage on Top Impact Limiter ............................................................................... 3-34 Figure 3.4.1-3 Temperature Contour Plot of Inner Shell Pre-Fire ConditionHAC Pin Damage on Top Impact Limiter ............................................................................... 3-35 Figure 3.4.1-4 Temperature Contour Plot of Package at the End of FireHAC Pin Damage on Top Impact Limiter .............................................................................................. 3-36 Figure 3.4.1-5 Temperature Contour Plot of Cask Body at the End of FireHAC Pin Damage on Top Impact Limiter ............................................................................... 3-37 Figure 3.4.1-6 Temperature Contour Plot of Package after Cool-DownHAC Pin Damage on Top Impact Limiter .............................................................................................. 3-38 Figure 3.4.1-7 Temperature Contour Plot of Cask Body after Cool-DownHAC Pin Damage on Top Impact Limiter ............................................................................... 3-39 Figure 3.4.1-8 Time-History Plot of Critical Package ComponentsHAC Pin Damage on Top Impact Limiter ................................................................................................... 3-40 Figure 3.4.1-9 Time-History Enhanced View Plot of Critical Package ComponentsHAC Pin Damage on Top Impact Limiter ......................................................................... 3-41 Figure 3.4.1-10 Maximum Temperature of the Inner ShellHAC Pin Damage on Top Impact Limiter ...................................................................................................................... 3-42 Figure 3.4.1-11 Maximum Temperature of Lead ShieldingHAC Pin Damage on Top Impact Limiter ...................................................................................................................... 3-43 Figure 3.4.2-1 Cask Model-HAC Pin Damage on Cask Body Side ................................................. 3-45 Figure 3.4.2-2 Temperature Contour Plot of Package Pre-Fire ConditionHAC Pin Damage on Cask Body Side ................................................................................................... 3-46 Figure 3.4.2-3 Temperature Contour Plot of Cask Body Pre-Fire ConditionHAC Pin Damage on Cask Body Side ..................................................................................... 3-47 Figure 3.4.2-4 Temperature Contour Plot of Inner Shell Pre-Fire ConditionHAC Pin Damage on Cask Body Side ..................................................................................... 3-48 Robatel Technologies, LLC Page TOC-19

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.4.2-5 Temperature Contour Plot of Package at the End of FireHAC Pin Damage on Cask Body Side ................................................................................................... 3-49 Figure 3.4.2-6 Temperature Contour Plot of Cask Body at the End of FireHAC Pin Damage on Cask Body Side ..................................................................................... 3-50 Figure 3.4.2-7 Temperature Contour Plot of Package after Cool-DownHAC Pin Damage on Cask Body Side ................................................................................................... 3-51 Figure 3.4.2-8 Temperature Contour Plot of Cask Body After Cool-DownHAC Pin Damage on Cask Body Side ..................................................................................... 3-52 Figure 3.4.2-9 Time-History Plot of Critical Package ComponentsHAC Pin Damage on Cask Body Side ........................................................................................................ 3-53 Figure 3.4.2-10 Time-History Enhanced View Plot of Critical Package ComponentsHAC Pin Damage on Cask Body Side ............................................................................... 3-54 Figure 3.4.2-11 Maximum Temperature of the Inner ShellHAC Pin Damage on Cask Body Side ........................................................................................................................... 3-55 Figure 3.4.2-12 Maximum Temperature of Lead ShieldingHAC Pin Damage on Cask Body Side ........................................................................................................................... 3-56 Figure 4-1 Information Flow for the Containment Review .......................................................... 4-2 Figure 4.1.2-1 Illustration of Containment Boundary ........................................................................ 4-4 Figure 4.3.1-1 Allowable Air/Helium Mixture Test Leakage Rates ................................................ 4-12 Figure 4.3.1-2 Allowable Helium Test Leakage Rates .................................................................... 4-13 Figure 4.4.4-1 Package Loading Curve for Hydrogen Generation - Decay Heat Limit Versus Waste Volume .......................................................................................................... 4-32 Figure 5-1 Information Flow for the Containment Review .......................................................... 5-4 Figure 5.3.1-1 NCT Model 1 & 2 ..................................................................................................... 5-11 Figure 5.3.1-2 NCT Model Tally Surfaces for Dose Rate Response Estimation .............................. 5-12 Figure 5.3.1-3 HAC Model 1 ........................................................................................................... 5-13 Figure 5.3.1-4 HAC Model 2 ........................................................................................................... 5-14 Figure 5.3.1-5 HAC Model General Tally Surfaces ......................................................................... 5-15 Figure 5.3.1-6 HAC Model Surface Tally ........................................................................................ 5-16 Figure 5.4.1-1 Fe-59 Spectrum Grouped into Generic Energy Lines .............................................. 5-21 Figure 5.4.4-1 Fluctuation in Radial Dose Rates (Cs-137) .............................................................. 5-28 Figure 5.4.4-2 Summary of Calculated Dose Rate Margins ............................................................ 5-31 Figure 5.4.4-3 Example of Media Density Effect ............................................................................ 5-33 Figure 5.4.4-4 NCT Maximum Gamma Dose Rates for Co-60 Content at 1.00 g/cc ...................... 5-35 Robatel Technologies, LLC Page TOC-20

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 5.5.5-1 Mass Restricted Filters - Radial Content Variation (500 lb cases shown) ............... 5-52 Figure 5.5.5-2 Mass Restricted Filters - Axial Content Variation (Top / 500 lb cases shown) ....... 5-53 Figure 5.5.5-3 Mass Restricted Filters - Radial Tally Locations...................................................... 5-54 Figure 5.5.5-4 Mass Restricted Filters - Axial Tally Locations ....................................................... 5-54 Figure 5.5.6-1 Cs-137Mesh Tally - NCT Side (500 lbs - Case 1) .................................................. 5-59 Figure 5.5.6-2 Co-60 Mesh Tally - NCT Side (500 lbs - Case 7) ................................................... 5-59 Figure 5.6.4-1 Low Density Hardware Mass Attenuation Coefficients ........................................... 5-64 Figure 5.6.4-2 High Density Hardware Mass Attenuation Coefficients .......................................... 5-64 Figure 5.6.4-3 Radial Content Variation (Low-density Hardware / 8,000 lb cases shown)............. 5-65 Figure 5.6.4-4 Axial Content Variation (Low-density Hardware / 8,000 lb cases shown) .............. 5-65 Figure 5.6.6-1 Mn-54 Mesh Tally - NCT Side (High-density / 1,000 lbs) ...................................... 5-76 Figure 5.6.6-2 Co-60 Mesh Tally - NCT Side (High-density / 1,000 lbs) ....................................... 5-76 Figure 7-1 Information Flow for the Operating Procedures Review ............................................ 7-2 Figure 7.1.1-1 Preparation for Loading Process Flowchart................................................................ 7-4 Figure 7.1.2-1 Loading of the RT-100 Process Flowchart ................................................................. 7-7 Figure 7.1.3-1 Preparation for Transport Process Flowchart ........................................................... 7-10 Figure 7.4.2-1 Lifting Yoke Arm Positioned on Cask ..................................................................... 7-15 Figure 7.4.2-2 Lifting Yoke Connections......................................................................................... 7-15 Figure 7.4.2-3 Lifting Yoke Secured with Locking Pin ................................................................... 7-15 Figure 7.4.2-4 Assembled Cask Ready to Lift ................................................................................. 7-15 Figure 7.4.4-1 Example Trailer Illustration ...................................................................................... 7-17 Figure 7.4.4-2 Loading of the RT-100 on Transportation Trailer .................................................... 7-17 Figure 7.5.1-1 Package Loading Curve for Hydrogen Generation - Decay Heat Limit Versus Waste Volume .......................................................................................................... 7-22 Figure 7.6.1-1 Mixed Shipment Loading Curve (Simplified Model)............................................... 7-42 Figure 8-1 Information Flow for the Acceptance Tests and Maintenance Program Review ....... 8-1 Figure 8.1.4-1 Cask Body Containment Boundary Test Apparatus ................................................... 8-5 Figure 8.1.4-2 Primary Lid Assembly Containment Boundary Test Apparatus ................................ 8-6 Figure 8.1.4-3 Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve ........................................................................................................ 8-9 Figure 8.1.4-4 Alternate Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve ........................................................................................... 8-10 Figure 8.1.4-5 Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve Cover Plate .................................................................................. 8-11 Robatel Technologies, LLC Page TOC-21

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 8.1.4-6 Alternate Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve Cover Plate ....................................................................... 8-11 Figure 8.3.2-1 Lead Solidification Diagram..................................................................................... 8-32 Figure 8.3.2-2 Effect of temperature on the physical properties of AISI 301 stainless steel

[Ref. 11] .................................................................................................................... 8-34 Robatel Technologies, LLC Page TOC-22

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 List of Tables Table 2.1.2-1 Load Combinations for RT-100 Cask Body Analyses ............................................... 2-5 Table 2.1.2-2 Structural Design Criteria for RT-100 ........................................................................ 2-5 Table 2.1.3-1 Assembly Weights and Center of Gravity Locations .................................................. 2-7 Table 2.2.1-1 Cask Temperature-Dependent Material Properties..................................................... 2-9 Table 2.2.1-2 Cask Temperature-Independent Material Properties................................................. 2-10 Table 2.2.1-3 Allowable Stresses for Cask Body Materials ........................................................... 2-10 Table 2.5.1-1 Summary of Results for Lifting Assembled Cask .................................................... 2-21 Table 2.5.2-1 Tie-down Arms Horizontal Angles .......................................................................... 2-29 Table 2.5.2-2 Calculated Values for Tie-Down Arms .................................................................... 2-32 Table 2.5.2-3 Calculated Forces for Tie-Down Arms..................................................................... 2-32 Table 2.6.7-1 NCT Side Drop Stress Summary .............................................................................. 2-53 Table 2.6.7-2 NCT End Drop Stress Summary............................................................................... 2-61 Table 2.7.1-1 Deceleration Loadings in RT-100 Cask Body Finite Element Analyses .................. 2-69 Table 2.7.1-2 HAC End Drop Stress Summary .............................................................................. 2-73 Table 2.7.1-3 HAC Side Drop Stress Summary ............................................................................. 2-82 Table 2.7.1-4 Corner Drop Component Accelerations ................................................................... 2-89 Table 2.7.1-5 HAC Corner Drop Stress Summary ......................................................................... 2-90 Table 2.7.3-1 HAC Pin Puncture Stress Summary ......................................................................... 2-94 Table 2.7.4-1 HAC Pressure Stress Summary .............................................................................. 2-103 Table 2.12.3-1 Foam Crush Strength Coefficients.......................................................................... 2-112 Table 2.12.4-1 Maximum Decelerations Summary for 9.0 m End-Drop Case ............................... 2-127 Table 2.12.4-2 Maximum Crush Depths for 9.0 m End-Drop Case ............................................... 2-127 Table 2.12.4-3 Maximum Decelerations Summary for 9.0 m Side-Drop Case .............................. 2-130 Table 2.12.4-4 Maximum Crush Depths for 9.0 m Side-Drop Case............................................... 2-130 Table 2.12.4-5 Maximum Decelerations for 9.0 m Corner-Drop Case........................................... 2-133 Table 2.12.4-6 Maximum Crush Depths for 9.0 m Corner-Drop Case........................................... 2-133 Table 2.12.4-7 Maximum Decelerations for 0.3 m End-drop Case ................................................ 2-137 Table 2.12.4-8 Maximum Crush Depths for 0.3 m End-drop Case ................................................ 2-137 Table 2.12.4-9 Maximum Decelerations for 0.3 m Side-drop Case ............................................... 2-140 Table 2.12.4-10 Maximum Crush Depths for 0.3 m Side-Drop Case............................................... 2-140 Table 2.12.4-11 Maximum Decelerations for 0.3 m Corner-Drop Case........................................... 2-143 Table 2.12.4-12 Maximum Crush Depths for 0.3 m Corner-Drop Case........................................... 2-143 Table 2.12.4-13 Actual Test Conditions - 3/10 Scale Cask Drop Tests............................................ 2-145 Robatel Technologies, LLC Page TOC-23

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.12.4-14 Foam Densities - 3/10 Scale Cask Drop Tests........................................................ 2-145 Table 2.12.4-15 Predicted and Actual Deceleration and Crush Depth - 3/10 Scale Cask End-Drop ........................................................................................................................ 2-146 Table 2.12.4-16 Predicted and Actual Deceleration and Crush Depth - 3/10 Scale Cask Corner-Drop ........................................................................................................................ 2-148 Table 2.12.4-17 Predicted and Actual Deceleration and Crush Depth - 3/10 Scale Cask Side-Drop ........................................................................................................................ 2-150 Table 2.12.5-1 Comparison of 3/10th Scale Drop Test Results and Analytical Method ................ 2-152 Table 2.12.6-1 Design Decelerations for Subsequent Cask Body Analysis ................................... 2-153 Table 2.13.2-1 Closure Bolt Loads for 9.0 m Corner-Drop ............................................................ 2-179 Table 2.13.3-1 Primary Lid Bolt Load Summary ........................................................................... 2-192 Table 2.13.3-2 Secondary Lid Bolt Load Summary ....................................................................... 2-192 Table 2.14.2-1 Lead Shrinkage Evaluation Results ........................................................................ 2-219 Table 2.15.4-1 Stress Concentration Factors .................................................................................. 2-222 Table 2.15.4-2 Sealing Surface Stress Summary ............................................................................ 2-222 Table 2.15.4-3 Lid Seal Groove Region Stresses ........................................................................... 2-222 Table 2.15.4-4 HAC Seal Region Displacement ............................................................................ 2-223 Table 3.1.3-1 RT-100 Maximum Normal Condition Temperature Summary .................................. 3-6 Table 3.1.3-2 RT-100 Maximum Calculated Temperature of Cask under HAC with Pin Puncture Damage on Top Impact Limiter .................................................................. 3-6 Table 3.1.3-3 RT-100 Maximum Calculated Temperature of Cask under HAC with Pin Puncture Damage at the Side of the Cask Body ......................................................... 3-7 Table 3.1.4-1 RT-100 Summary of Maximum Normal and Hypothetical Accident Condition Pressures ..................................................................................................................... 3-7 Table 3.2.1-1 Temperature-Independent Material Properties ....................................................... 3-9 Table 3.2.1-2 Temperature-Dependent Material PropertiesStainless Steel 304.......................... 3-10 Table 3.2.1-3 Temperature-dependent Material PropertiesLead ................................................. 3-11 Table 3.2.1-4 Temperature-dependent Material PropertiesCeramic Paper ................................. 3-11 Table 3.2.2-1 Component Specifications - Minimum and Maximum Temperatures ..................... 3-12 Table 3.2.3-1 Maximum Temperature Limits for RT-100 Content Materials ................................ 3-13 Table 4.1.4-1 Bolt Torque Requirements.......................................................................................... 4-6 Table 4.1.5-1 Cask Cavity Dimensions ............................................................................................ 4-6 Table 4.1.5-2 Cask Cavity Volume ................................................................................................... 4-6 Table 4.1.5-3 Parameters for Normal Transport and Accident Conditions ...................................... 4-7 Robatel Technologies, LLC Page TOC-24

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.1.5-1 Leakage Tests of the RT-100 Package ....................................................................... 4-9 Table 4.3.1-1 Helium and Air Viscosity ......................................................................................... 4-11 Table 4.3.1-2 Allowable Helium Test Leakage Rates, cm³/sec ...................................................... 4-14 Table 4.4.1-1 G Values (Molecules/100eV) for Potential Content Materials ................................. 4-18 Table 4.4.1-2 Effective G Values (Molecules/100eV) for Potential Content Materials ................. 4-19 Table 4.4.1-3 Activation Energy ..................................................................................................... 4-20 Table 4.4.1-4 Bounding G Values for Contents at Maximum NCT Temperature .......................... 4-21 Table 4.4.3-1 Effective G Values and Corresponding Values for Contents ................................ 4-26 Table 4.4.3-2 Secondary Container Volumes and Allowable Shoring Volume ............................. 4-30 Table 4.4.4-1 Conditions and Justifications for using Package Loading Curve (Figure 4.4.4-1) ......................................................................................................... 4-33 Table 4.4.5-1 Conditions for Shipper to use the Detailed Analysis ................................................ 4-35 Table 4.4.5-2 G-values and -Fractions for a Range of Alpha/Gamma Decay Heat Distributions ............................................................................................................. 4-36 Table 5.1.2-1 Resin/Filter Summary Table of External Radiation Levels (Exclusive Use) ............. 5-6 Table 5.2.1-1 One Curie Co-60 Gamma Source Term ..................................................................... 5-7 Table 5.3-1 Model Shielding Thicknesses ..................................................................................... 5-8 Table 5.3.2-1 RT-100 Material Composition Summary ................................................................. 5-18 Table 5.4.3-1 ANSI/ANS 6.1.1-1977 - Gamma Flux-to-Dose Conversion Factors [Ref. 13] ....... 5-25 Table 5.4.4-1 Maximum Dose Rates and Responsible Radionuclides ........................................... 5-26 Table 5.4.4-2 Media Composition Comparison .............................................................................. 5-29 Table 5.4.4-3 Media Density Comparison ...................................................................................... 5-30 Table 5.4.4-4 NCT Dose Rate Responses Due to Bremsstrahlung ................................................. 5-32 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading ........................ 5-36 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading ....................... 5-43 Table 5.5.1-1 Co-60 Mass Restricted Filter Maximum Radiation Level Summary ....................... 5-50 Table 5.5.6-1 Ag-110m Dose Rates and Activity Limits................................................................ 5-56 Table 5.5.6-2 Cs-137 Dose Rates and Activity Limits ................................................................... 5-56 Table 5.5.6-3 Co-58 Dose Rates and Activity Limits ..................................................................... 5-56 Table 5.5.6-4 Co-60 Dose Rates and Activity Limits ..................................................................... 5-57 Table 5.5.6-5 Cs-134 Dose Rates and Activity Limits ................................................................... 5-57 Table 5.5.6-6 Fe-59 Dose Rates and Activity Limits...................................................................... 5-57 Table 5.5.6-7 Mn-54 Dose Rates and Activity Limits .................................................................... 5-58 Table 5.5.6-8 Zn-65 Dose Rates and Activity Limits ..................................................................... 5-58 Robatel Technologies, LLC Page TOC-25

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.5.7-1 Mass Restricted Filter Specific Activity Limit Summary ........................................ 5-61 Table 5.6.1-1 Co-60 Activated Hardware Maximum Radiation Level Summary .......................... 5-62 Table 5.6.6-1 RT-100 Activated Hardware Dose Rate Summary - 1,000 lbs ................................ 5-67 Table 5.6.6-2 RT-100 Activated Hardware Dose Rate Summary - 2,000 lbs ................................ 5-68 Table 5.6.6-3 RT-100 Activated Hardware Dose Rate Summary - 8,000 lbs ................................ 5-69 Table 5.6.6-4 RT-100 Activated Hardware Dose Rate Summary - 13,000 lbs .............................. 5-70 Table 5.6.6-5 Ag-110m Activated Hardware Specific Activity Results ......................................... 5-71 Table 5.6.6-6 Cs-137 Activated Hardware Specific Activity Results............................................. 5-72 Table 5.6.6-7 Co-58 Activated Hardware Specific Activity Results .............................................. 5-72 Table 5.6.6-8 Co-60 Activated Hardware Specific Activity Results .............................................. 5-73 Table 5.6.6-9 Cs-134 Activated Hardware Specific Activity Results............................................. 5-73 Table 5.6.6-10 Fe-59 Activated Hardware Specific Activity Results ............................................... 5-74 Table 5.6.6-11 Mn-54 Activated Hardware Specific Activity Results ............................................. 5-74 Table 5.6.6-12 Zn-65 Activated Hardware Specific Activity Results .............................................. 5-75 Table 5.6.7-1 Activate Hardware Specific Activity Limit Summary ............................................. 5-78 Table 5.7.1-1 List of Gamma Radionuclides with Greater Than 1 Day Half Life .......................... 5-79 Table 5.7.2-1 NCT Gamma Dose Rate Responses (mrem/hr/Ci) ................................................... 5-81 Table 5.7.2-2 HAC Gamma Dose Rate Responses (mrem/hr/Ci)................................................... 5-89 Table 5.7.3-1 Radionuclide Maximum Ci/g Loading Limits based on Gamma Response ............. 5-97 Table 7.4.5-1 Lid Bolt Tightening Torques .................................................................................... 7-18 Table 7.4.5-2 Tightening Torques - Other Parts ............................................................................. 7-18 Table 7.5.1-1 Conditions for using Package Loading Curve (Excerpt from Table 4.4.4-1) ........... 7-22 Table 7.5.1-2 Secondary Container and Allowable Shoring Volumes (Excerpt from Table 4.4.3-2)...................................................................................................................... 7-23 Table 7.5.2-1 Conditions for Shipper to use the Detailed Analysis (From Table 4.4.5-1) ............. 7-25 Table 7.5.2-2 G-values and -Fractions for a Range of Alpha/Gamma Decay Heat Distributions (Excerpt from Table 4.4.5-2) .............................................................. 7-26 Table 7.6.1-1 RT-100 Loading Table Illustration - Filter and Resin Sheet.................................... 7-33 Table 7.6.1-2 RT-100 Loading Table Illustration - Low-density Hardware Sheet ........................ 7-33 Table 7.6.1-3 RT-100 Loading Table Illustration - High-density Hardware Sheet........................ 7-33 Table 7.6.1-4 RT-100 Loading Table Illustration - Loading Summary Sheet ............................... 7-34 Table 7.6.1-5 Turkey Point Loading Table Example ...................................................................... 7-38 Table 7.6.1-6 Turkey Point Example Loading Summary ............................................................... 7-38 Table 7.6.1-7 St. Lucie Loading Table Example ............................................................................ 7-40 Robatel Technologies, LLC Page TOC-26

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-8 St. Lucie Example Loading Summary ...................................................................... 7-40 Table 7.6.1-9 Mixed Loading Example (resin/filter sheet) ............................................................. 7-41 Table 7.6.1-10 Mixed Loading Example (high-density hardware sheet) .......................................... 7-41 Table 7.6.1-11 Mixed Loading Example (loading summary sheet) .................................................. 7-42 Table 7.6.1-12 Maximum Co-60 Filter and Resin Loading (<500 lbs) ............................................ 7-43 Table 7.6.1-13 Maximum Co-60 Filter and Resin Loading (<1,000 lbs) ......................................... 7-43 Table 7.6.1-14 Maximum Co-60 Filter and Resin Loading (<1,500 lbs) ......................................... 7-43 Table 7.6.1-15 Maximum Co-60 Filter and Resin Loading (>1,500 lbs) ......................................... 7-43 Table 7.6.1-16 Failed Loading Table Example................................................................................. 7-44 Table 7.6.1-17 Failed Loading Table Example (Loading Summary) ............................................... 7-44 Table 7.6.1-18 Radionuclide Activity Concentration Limits (8 Explicitly Analyzed) ..................... 7-45 Table 7.6.1-19 Radionuclide Activity Concentration Limits (Generic Energy Line Method) ......... 7-46 Table 8.1.5-1 Material Specifications for O-Rings ......................................................................... 8-12 Table 8.1.5-2 Basic Requirements for O-Rings .............................................................................. 8-13 Table 8.1.5-3 Supplementary Requirements for O-Rings ............................................................... 8-13 Table 8.1.5-4 Critical Characteristics of Ceramic Paper................................................................. 8-14 Table 8.1.5-5 Materials and Reference Standards for Carbon Steel and Alloy Steel Fasteners ..... 8-15 Table 8.1.5-6 Chemical Composition for Carbon Steel and Alloy Steel Fasteners ........................ 8-16 Table 8.1.5-7 Mechanical Properties of Carbon Steel and Alloy Steel Fasteners........................... 8-17 Table 8.1.5-8 Materials and Reference Standards for Stainless Steel Fasteners ............................. 8-18 Table 8.1.5-9 Chemical Composition for Stainless Steel Fasteners ............................................... 8-18 Table 8.1.5-10 Mechanical Properties of Stainless Steel Fasteners .................................................. 8-18 Table 8.1.5-11 Chemical Composition for Stainless Steel Bars ....................................................... 8-19 Table 8.1.5-12 Mechanical Properties of Stainless Steel Bars.......................................................... 8-19 Table 8.1.5-13 Chemical Composition for Threaded Inserts ............................................................ 8-20 Table 8.1.5-14 Mechanical Properties of Threaded Inserts .............................................................. 8-20 Table 8.2.2-1 Volume of the Interspaces between the O-rings ....................................................... 8-23 Table 8.3.1-1 RT-100 Leakage Test Types..................................................................................... 8-30 Table 8.3.1-2 Allowable Helium Leakage Rates ............................................................................ 8-30 Robatel Technologies, LLC Page TOC-27

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 List of Attachments .4-1 RT100 NM 1000 Rev. G Bill of Material............................................................ 1-13 .4-2 RT100 PE 1001-1 Rev. H Robatel Transport Package RT-100 General Assembly Sheet 1/2 .................................................................................................. 1-19 .4-3 RT100 PE 1001-2 Rev. H Robatel Transport Package RT-100 General Assembly Sheet 2/2 .................................................................................................. 1-20 .4-4 RT100 PRS 1011 Rev. E Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Cask Body .............................................................................. 1-21 .4-5 RT100 PRS 1013 Rev. C Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Secondary Lid ........................................................................ 1-22 .4-6 RT100 PRS 1031 Rev. D Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Lower Impact Limiter ............................................................ 1-23 .4-7 RT100 PRS 1032 Rev. D Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Upper Impact Limiter ............................................................ 1-24 .4-8 102885 MD 1031-06 Rev. F Robatel Transport Package RT-100 Sub Assembly Fabrication Drawing Impact Limiter Foam ............................................. 1-25 .4-9 RT-100 Cask as Prepared for Transport with Approximate Trailer Dimensions ..... 1-26 .12-1 General Plastics Foam Product Information Sheets ............................................... 2-154 .5-1 EPDM Temperature Specifications .......................................................................... 3-61 .5-2 Seal Material EPDM Working Temperature ............................................................ 3-62 .5-3 Water Vapor Pressure Reference (80°C) .................................................................. 3-63 .5-4 Water Vapor Pressure Reference (150°C) ................................................................ 3-66 .5-1 EPDM Temperature Specifications .......................................................................... 4-37 .5-2 Seal Material EPDM Working Temperature ............................................................ 4-38 .5-3 Seal Material EPDM Helium gas permeation rate.................................................... 4-39 .5-4 Seal Material EPDM Characteristics With Respect to Damage by Radiation and Hardness Concerns ............................................................................................ 4-41 .5-5 Additional Support Information about EPDM Resistance to Radiation Up to 5x108 Rads While Retaining Reasonable Flexibility and Strength, Hardness and Very Good Compression Set Resistance ........................................................... 4-45 Robatel Technologies, LLC Page TOC-28

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

1. GENERAL INFORMATION ROBATEL Technologies, LLC (RT) submits this Application and Safety Analysis Report (SAR),

Revision 10, to the Nuclear Regulatory Commission (NRC) to amend and renew the Certificate of Compliance (CoC) No. 9365, Revision No. 2 for the Model RT-100 Type B(U) Cask Package (RT-100).

This application is intended to meet the NRC regulation 10 CFR Part 71, and the Canadian Nuclear Safety Commission (CNSC) regulation SOR/2000-208, and was prepared following the guidance of NUREG 1886. The RT-100 package and the SAR meet the more stringent of the two codes.

Chapter 1 of the SAR provides General Information that feeds information to later sections in this application according to Figure 1-1 on the following page. The RT-100 meets the following general requirements for all packages:

o The smallest overall dimension of the RT-100 is not less than 10 cm (4 in.).

o The outside of the RT-100 incorporates a feature that, while intact, is evidence that the package has not been opened by unauthorized persons.

1.1 Introduction The purpose of this application is to request the amendment of the CoC for the Model No. RT-100 type B(U) cask to allow for the transport of activated hardware or activated metal (terms used interchangeably) contents packaged in a secondary container in addition to contaminated spent resins and filters as specified on the current CoC. Moreover, this license amendment is to provide the flexibility to ship contaminated spent resins/filters and activated hardware of varying activities by limiting the wastes mass.

This application does not request the packaging and/or transport of fissile material in quantities exceeding those exempted from consideration in accordance with 10 CFR 71.15 [Ref. 2] and thus, the Criticality Safety Index (CSI) is non-applicable.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 1-1 Information Flow for General Information General Information Review Introduction Package Description General Requirements

Purpose

Packaging

Minimum Size

Summary Information

Contents

Evidence of Unauthorized Opening

Containment Boundary

Operational Features

Drawings Structural Thermal Containment Shielding Evaluation Evaluation Evaluation Evaluation

Package Category

Dimensions

Dimensions

Dimensions

Materials

Materials

Contents

Materials

Dimensions

Decay Heat

Materials

Contents

Weights

Containment

Exclusive /

Boundary Nonexclusive use Criticality Operating Acceptance Tests Evaluation Procedures and Maintenance

Not Applicable

Operational

Codes and Features Standards

General Restrictions

Dimensions and

Tamper-Indicating Tolerances Device

Materials

Contents Robatel Technologies, LLC Page 1-2

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.2 Package Description Section 1.2 provides a summary of all design aspects of the RT-100. A general arrangement of the RT-100 cask is included in Appendix 1.4. The general arrangement depicts the package dimensions and the materials of construction. Figure 1.2.1-1 shows the major components of the RT-100 as an exploded artist view with the various components labeled.

1.2.1 Packaging Section 1.2.1 provides details regarding overall dimensions, weight, containment, shielding, criticality, structural features, heat transfer features and package markings.

1.2.1.1 Overall Dimensions The package consists of a stainless-steel and lead cylindrical shipping cask with a pair of cylindrical foam-filled impact limiters installed on each end. The package Figure 1.2.1-1 RT-100 Cask Package configuration is shown in Figure 1.2.1-1. Artist Concept The internal cavity dimensions are 1730 mm in diameter and 1956 mm high. The cylindrical cask body is comprised of a 35 mm thick outer stainless-steel shell and a 30 mm thick inner stainless-steel plate. The annular space between the shells is filled with 90 mm thick lead.

The base of the cask consists of a 30 mm thick stainless steel outer bottom plate, a 75 mm thick gamma shield of poured lead, and a 50 mm thick stainless steel inner bottom forging.

The primary lid consists of a 210 mm thick stainless steel forging. The primary lid is fastened to the cask body with thirty-two (32) M48 hex head bolts.

The secondary lid is made of 100 mm thick stainless steel plate, a 60 mm thick lead gamma shield and a 10 mm thick stainless steel plate. The secondary lid is attached to the primary lid with eighteen (18) M36 hex head bolts.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.2.1.2 Weight The maximum gross weight of the RT-100 including impact limiters is 41,500 kg (including the maximum payload weight of 6,804 kg). The maximum (empty) weight of the RT-100 including impact limiters is 34,696 kg.

1.2.1.3 Containment Features The containment vessel of the RT-100 cask consists of the inner shell, the bottom forging, the top flange, the primary lid, the primary lid inner O-ring, the stainless steel vent port cover plate and its inner O-ring, the secondary lid and the secondary lid inner O-ring. The containment system prevents leakage of radioactive material from the cask cavity and allows pre-shipment leakage testing of the assembled cask configuration.

1.2.1.4 Neutron and Gamma Shielding Features The RT-100 is not designed to carry fissile material or neutron sources (except typical small quantities consistent with contaminated resins and filters as discussed in Chapter 5) and thus, provision of neutron shielding is not required for the RT-100.

In regards to gamma shielding, the RT-100 cask walls provide a shield thickness of 90 mm of lead and 70 mm of stainless steel including the thermal shield plate of 5 mm thickness (65 mm used for HAC analysis). The cask bottom end provides a shield thickness of 75 mm of lead and 80 mm of stainless steel. The top end provides a shield thickness of 210 mm of stainless steel for the primary lid and a shield thickness of 60 mm of lead and 110 mm of stainless steel for the secondary lid.

Contents are limited such that the radiological shielding provided assures compliance with U.S.

Department of Transportation (DOT) regulatory requirements.

1.2.1.5 Shielding Features for Personnel Barriers The RT-100 does not require the use of personnel barriers to meet 10 CFR 71 dose rate limits.

1.2.1.6 Criticality Control Features The RT-100 contents are activated hardware and spent resins and filters from commercial nuclear power plants that contain only trace quantities of fissile radionuclides. As such, the contents meet the requirements of 10 CFR 71.15 [Ref. 2] and are exempt from classification as fissile material.

As a result, the RT-100 does not require any criticality control features.

1.2.1.7 Structural Features - Lifting and Tie-Down Devices The RT-100 cask employs lifting devices that are a structural part of the package. Two lifting pockets are welded to the cylindrical cask body as shown in Drawing RT100 PE 1001-02, Rev. H (Chapter 1, Appendix 1.4, Attachment 1.4-3). The pockets engage the arms of a separate lifting yoke used to lift the package. When not in use for package lifting, the pockets are rendered inoperable so they cannot be inadvertently used as cask tie-downs. Removable lifting lugs are Robatel Technologies, LLC Page 1-4

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 utilized for removal and handling of the primary and secondary lids, as well as the impact limiters.

Refer to Chapter 2, Section 2.5.1 for a detailed analysis of the structural integrity of the lifting devices.

Two tie-down arms are welded to the external cask shell and are considered a structural part of the package. When not in use for package tie-down, the arms' holes are rendered inoperable preventing the tie-down arms from being used to lift the packaging. Refer to Chapter 2, Section 2.5.2 for a detailed analysis of the structural integrity of the tie-down arms.

1.2.1.8 Structural Features - Impact Limiters The impact limiters have an outside diameter of 2587 mm. The lower impact limiter extends 494 mm beyond the base of the cask. The upper impact limiter extends 498 mm beyond the cask primary lid. The impact limiter external shells are stainless-steel, allowing them to withstand large plastic deformation without fracturing. The volume inside the shell is filled with crushable shock-absorbing and thermal-insulating polyurethane foam. The polyurethane is preformed and inserted into the shell to the void space. The use of preformed foam ensures homogeneous density. Several different foam densities are used to customize the shock absorbing performance of the impact limiters during hypothetical accident conditions. The rationale for use of preformed foam blocks and the use of different foam densities is presented in detail in Chapter 2, Section 2.2.

The impact limiters are attached to the cask via two stainless-steel bolt ring flanges located on the exterior cask body. The flanges are welded along the cask circumference and considered a structural part of the package. Each impact limiter is equipped with twelve (12) M36 studs and attached to the bolt ring using twelve (12) M36 stainless steel hex head nuts. The purpose of the bolt rings and bolts are to ensure the impact limiters remain attached to the cask body for all Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) events.

Additionally, use of bolt rings facilitates removal of the impact limiters during loading and unloading operations.

1.2.1.9 Structural Features - Internal Supporting or Positioning Features The RT-100 cask interior has no supporting or positioning features. The waste contents shall be pre-packaged in liners and placed into the cask cavity. Waste liners may require appropriate shoring to prevent movement during transit. It is the responsibility of the shipper to provide shoring that meets DOT requirements.

1.2.1.10 Structural Features - Outer Shell or Outer Packaging The external surface of the cylindrical cask body is comprised of a 35 mm thick stainless-steel outer shell.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.2.1.11 Structural Features - Packaging Closure Device The main packaging closure device is the primary lid that consists of a 210 mm thick stainless steel forging as described in Section 1.2.1.1. The primary lid is fastened to the cask body with thirty-two (32) M48 hex head bolts.

The secondary lid also represents a closure device for the cask and is made of 100 mm thick stainless steel plate with lead shielding and another stainless steel plate as described in Section 1.2.1.1. The secondary lid is attached to the primary lid with eighteen (18) M36 hex head bolts.

1.2.1.12 Structural Features - Heat Transfer Features The RT-100 relies on the insulating properties of the impact limiter polyurethane foam and the cask body ceramic fiber thermal shield to minimize heat input during the hypothetical fire accident event. See Chapter 3, Section 3.4 for details.

There are no special features designed to dissipate heat from the cask.

1.2.1.13 Structural Features - Packaging Markings The side of the cask body is marked with the Model Number of the cask RT-100, the Certificate of Compliance No., Empty Weight, Type B(U)-96, UN 2916 and other required data.

1.2.1.14 Additional Information o RT-100 cask has one configuration as depicted in the engineering drawings provided in Appendix 1.4, Attachments 1.4-1 thru 1.4-8.

o The RT-100 has no receptacles.

o Pressure test ports are provided between the twin O-rings for the primary lid, between the O-rings for the secondary lid, and between the O-rings for the vent port cover plate.

These ports facilitate leak testing of the package in accordance with ANSI N14.5-2014

[Ref. 4].

o The vent port is provided for venting pressures within the containment cavity which may be generated during transport and prior to lid removal. Each port is sealed with an EPDM O-ring. Specification information for all O-rings is contained in Chapter 4, Section 4.1.3.

o The RT-100 does not rely on any coolants to perform its function of providing safe transportation of its radioactive contents.

o There are no external/internal protrusions other than the tie-down arms previously described.

1.2.2 Contents The authorized contents of the RT-100 are generally described in Section 1.2.2. The radioactive contents are described to the extent required to demonstrate compliance with 10 CFR 71 Robatel Technologies, LLC Page 1-6

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 requirements relating to the structural, thermal and shielding performance of the cask.

1.2.2.1 Identification and Maximum Quantity of Radioactive Material The contents of the RT-100 cask are limited to activated hardware and contaminated resins and filters containing byproduct or otherwise radioactive nuclear material.

The maximum quantity of material is defined as a Type B quantity of radioactive materials not to exceed 3000 A2. The activity of beta, gamma and neutron emitting radionuclides will not exceed the limits established in the shielding evaluation provided in Chapter 5 and using the procedure presented in Chapter 7.

1.2.2.2 Identification and Maximum Quantity of Fissile Material The RT-100 will not transport fissile material exceeding the quantities exempt in 10 CFR 71.15

[Ref. 2]. Thus, Section 1.2.2.2 is non-applicable.

1.2.2.3 Physical and Chemical Form - Density, Moisture Content and Moderators The type/form of material is defined as byproduct, source, or special nuclear material in the form of resins, filters, activated low-density hardware, activated high-density hardware, and mixtures of resins, filters, low-density hardware, and high-density hardware contained within a secondary container. The chemical forms of the contents are resins and filter media containing radioactive materials and metallic activated hardware segments in the form of dispersible solids. There are no contents in powdered form. The resin and filter contents may include the metal housings associated with the media.

1.2.2.3.1 Ion-Exchange Resins Single or mixed bed ion exchange resins are used in deep bed filter demineralizers for reduction of particulate matter and dissolved contaminants in utility power plant condensates. Radioactive waste systems in nuclear power plants include ion exchange systems for the removal of trace quantities of radioactive nuclides from water that will be released to the environment. The primary resin system used is the mixed bed.

Conventional ion exchange resins consist of a cross-linked polymer matrix with a relatively uniform distribution of ion-active sites throughout the structure. Ion exchange resin materials are sold as spheres or sometimes granules with a specific size and uniformity to meet the needs of a particular application. Ion exchange resins can contain up to 66% water when delivered from the manufacturer. This is essentially the same moisture content within the resin when delivered for disposal. The majority are prepared in spherical (bead) form, either as conventional resin with a polydispersed particle size distribution from about 0.3 mm to 1.2 mm (50-16 mesh) or as uniform particle sized (UPS) resin with all beads in a narrow particle size range. In the water swollen state, ion exchange resins typically show a specific gravity of 1.1-1.5. The bulk density as installed in a Robatel Technologies, LLC Page 1-7

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 column includes a normal 35-40 percent voids volume for a spherical conventional resin product.

Bulk densities in the range of 560-960 g/l (35-60 lb/ft3) are typical for wet resinous products [Ref.

7].

The contents are limited by the maximum overall weight limit of 6,804 kg as described in Section 1.2.1.2. The radioactive inventory of the contents are limited as a function of the activity concentration as described in Chapter 5.

1.2.2.3.2 Filters Filters packaged in the secondary liner are designed for use in a nuclear power plants pure water chemistry; therefore, the housings are a non-corrosive and non-reactive material. Filter housings may be stainless steel or a thermoplastic such as polyethylene or polypropylene. They are designed to filter radioactive material from the water, and thus are acceptable for use in a radiation environment. The filter housings do not interact with the secondary container and therefore do not interact with the RT-100 metal cavity.

1.2.2.3.3 Activated Hardware Activated hardware contents include low-density hardware, such as aluminum and zircoloy and high-density hardware, such as steel and Inconel. Low-density hardware is limited to hardware with a density greater than or equal to 2 g/cm3 and less than 7.5 g/cm3. High density hardware is limited to hardware with a density greater than or equal to 7.5 g/cm3 and less than or equal to 9.0 g/cm3. Common examples of activated hardware that could be shipped in the RT-100 Cask include but are not limited to fuel channels, velocity limiters, and reactor vessel internals from PWRs and BWRs.

1.2.2.3.4 Secondary Containers Secondary containers may be constructed of carbon steel or stainless steel, or a thermoplastic such as polyethylene or polypropylene. The secondary containers are used to package contaminated spent resins/filters, activated low-density hardware, activated high-density hardware, or a mixture of spent resins/filters with activated hardware generated by nuclear power plants. There is a long history of transporting resins and filters via typical polyethylene or metal liners in metal casks by the nuclear power industry and other low-level waste generators. Secondary containers are required to be passively vented within the cask cavity during shipment. The RT-100 stainless steel inner cavity does not interact with polyethylene or metal liners typically used in the nuclear industry for the shipment of resins and filters. Secondary containers may be positioned or braced within the cavity using shoring. This shoring may be constructed of carbon steel or stainless steel, wood, or a thermoplastic material or any combination thereof.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 is placed between the secondary containers and the cask cavity liner to prevent movement during accident conditions. Providing appropriate shoring is the responsibility of the shipper.

1.2.2.5 Use of Non-Fissile Materials as Neutron Absorbers/Moderators The RT-100 does not contain non-fissile materials as neutron absorbers/moderators.

1.2.2.6 Chemical/Galvanic/Gas Generation Chemical Reaction and Galvanic Reactions The contents do not include materials that may cause any significant chemical, galvanic, or other reaction.

Gas Generation Secondary packages containing water and/or organic substances may generate combustible gases via radioanalytical reactions. A maximum molar quantity of 5% hydrogen by volume at standard temperature and pressure is allowed. The time duration is calculated as twice the expected shipment time.

Determination of hydrogen generation is made using the methods in NUREG/CR-6673 [Ref. 5],

Hydrogen Generation in TRU Waste Transportation Packages, and supplemented with data from EPRI NP-5977 [Ref. 6], Radwaste Radiolytic Gas Generation Literature Review. NUREG/CR-6673 provides equations that allow prediction of the hydrogen concentration as a function of time for simple nested enclosures and for packages containing multiple contents packaged within multiple nested confinement layers. The inputs to these equations include the bounding effective G(H2)-value for the contents, the G(H2)-values for the packaging material(s), the void volume in the containment vessel and in the confinement layers (when applicable), the temperature when the package was sealed, the temperature of the package during transport, and the contents decay heat.

EPRI NP-5799 provides G-Values for a wide range of ion exchange resins [Ref. 6].

For any package delivered to a carrier for transport, the secondary container is prepared for shipment in the same manner in which the determination for gas generation is made. Shipment period begins when the package is prepared (sealed) and is completed within a time period that is one half the time used in the hydrogen generation calculation. It is the shippers responsibility to ensure that hydrogen generation in the cavity will be below 5% by volume, representing the lower flammability limit for hydrogen. The maximum allowable shipping time is not restricted for any other reason. Detailed discussion of the hydrogen generation calculations are provided in Chapter 4, Section 4.4, and Chapter 7, Section 7.5.

Secondary packages with radioactive contents less than Low Specific Activity (LSA) and shipped within 10 days of preparation (or within 10 days of venting the secondary container) do not require Robatel Technologies, LLC Page 1-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 a determination of hydrogen gas generation or a restriction on shipping time. Activated hardware only shipments with no hydrogenous materials do not require hydrogen gas generation analysis.

1.2.2.7 Maximum Weight of Contents and Payload All contents shall be packaged in a secondary container (liner). The maximum gross weight of payload is 6,804 kg including the secondary container (liner) and shoring. The maximum payload of activated hardware contents is limited to 5,896 kg (~900 kg for secondary container and shoring).

1.2.2.8 Maximum Decay Heat The maximum decay heat of the RT-100 contents is 200 watts.

1.2.2.9 Loading Restrictions Contents that are prohibited include explosives, non-radioactive pyrophoric materials, and corrosives (pH less than 2 or greater than 12.5). Pyrophoric radionuclides may be present only in residual amounts less than 1% by weight. Materials that may auto-ignite or undergo phase transformation at temperatures less than 140 °C, with the exception of water, are not included in the contents. As required by 10 CFR 71.43(d) [Ref. 2], the contents do not include materials that may cause any significant chemical, galvanic, or other reactions.

1.2.2.10 Contents for the Certificate of Compliance The type and form of material is defined as byproduct, source, or special nuclear material in the form of dewatered or grossly dewatered spent resins/filters, activated hardware, or mixtures of resins/filters with activated hardware contained within secondary container(s). Secondary containers are required to be passively vented within the cask cavity during shipment. The maximum bulk density of resins and filters may not exceed 1.0 g/cm3. Activated hardware contents include low-density and high-density hardware. Low-density hardware (e.g., aluminum and zircoloy) is limited to hardware with a density greater than or equal to 2 g/cm3 and less than 7.5 g/cm3. High-density hardware (e.g., steel and Inconel) is limited to hardware with a density greater than or equal to 7.5 g/cm3 and less than or equal to 9.0 g/cm3. The maximum payload of resins/filters and activated hardware including contents, secondary containers, and shoring is limited to 6,804 kg. The maximum payload of activated hardware is limited to 5,896 kg. The maximum quantity of material is defined as a Type B quantity of radioactive materials not to exceed 3000 A2. The activity of alpha, beta, gamma and neutron emitting radionuclides does not exceed the limits established in the shielding evaluation provided in Chapter 5 and using the loading table provided in Appendix 7.6, Section 7.6.1. The contents may include fissile materials provided at least one of the paragraphs (a) through (f) of 10 CFR 71.15 [Ref. 2] is met.

1.2.3 Special Requirements for Plutonium The RT-100 will not contain plutonium in solid form. Therefore, the requirements of 10 CFR 71.63 Robatel Technologies, LLC Page 1-10

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[Ref. 2] specifying that more than 0.74 TBq (20 Ci) of plutonium must be in solid form do not apply.

1.2.4 Operational Features The RT-100 has no complex operational requirements. The various valves, connections, openings, seals and containment boundaries are depicted in the drawings provided in Appendix 1.4, Attachments 1.4-1 through 1.4-8. There are no piping systems associated with the RT-100 cask.

1.3 Engineering Drawings and Additional Information Appendix 1.4 contains the engineering drawings (Attachments 1.4-1 thru 1.4-8) and additional information associated with the RT-100.

1.3.1 Engineering Drawings The RT-100 drawings are enclosed in Appendix 1.4, Attachments 1.4-1 thru 1.4-8, and contain the following information:

o Safety features (primary and secondary lids, seals, bolts, containment boundary, and shielding) o Materials list, dimensions, vent and leak test ports and weld inspection requirements o Weld joint requirements o Details of gasket joints Appendix 1.4 does not include detailed construction drawings.

1.3.2 Conformance to Approved Design The RT-100 cask will be fabricated in accordance with the drawings referenced in the CoC.

1.3.3 Referenced Pages All referenced pages are generally available to the public.

1.3.4 Special Fabrication Procedures Fabrication of the RT-100 involves standard cask fabrication techniques.

1.3.5 Package Category The RT-100 is categorized as a Type B(U)-96 Package.

1.3.6 Supplemental Information This application contains no supplemental information.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.4 Appendix Appendix 1.4 contains Proprietary Information that Robatel requests be withheld from public disclosure under 10 CFR 2.390. This request is in accordance with the Robatel Affidavit and as requested in 10 CFR 2.390.

Attachment 1.4-1 RT100 NM 1000 Rev. G Bill of Material Attachment 1.4-2 RT100 PE 1001-1 Rev. H Robatel Transport Package RT-100 General Assembly Sheet 1/2 Attachment 1.4-3 RT100 PE 1001-2 Rev. H Robatel Transport Package RT-100 General Assembly Sheet 2/2 Attachment 1.4-4 RT100 PRS 1011 Rev. E Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Cask Body Attachment 1.4-5 RT100 PRS 1013 Rev. C Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Secondary Lid Attachment 1.4-6 RT100 PRS 1031 Rev. D Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Lower Impact Limiter Attachment 1.4-7 RT100 PRS 1032 Rev. D Robatel Transport Package RT-100 Cask Sub Assembly Weld Map Upper Impact Limiter Attachment 1.4-8 102885 MD 1031-06 Rev. F Robatel Transport Package RT-100 Sub Assembly Fabrication Drawing Impact Limiter Foam Attachment 1.4-9 RT-100 Cask as Prepared for Transport with Approximate Trailer Dimensions Robatel Technologies, LLC Page 1-12

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 1.4-9 RT-100 Cask as Prepared for Transport with Approximate Trailer Dimensions Robatel Technologies, LLC Page 1-26

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.5 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL 71.15 71.43(d) 71.63

3. Robatel Technologies, LLC Application and Safety Analysis Report, Revision 7, for the Model RT-100 Cask Package, dated December 05, 2018.

4. ANSI N14.5-2014, "American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

5. NUREG/CR-6673, "Hydrogen Generation in TRU Waste Transportation Packages,"

Anderson, B., Sheaffer, M., & Fischer, L., Lawrence Livermore National Laboratory, Livermore, CA, May 2000.

6. EPRI NP-5977, Radwaste Radiolytic Gas Generation Literature Review, Electric Power Research Institute, September 1988.

7. Resin and Filter Handbook - Primers and Product Information Robatel Technologies, LLC Page 1-27

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2. STRUCTURAL EVALUATION Chapter 2 describes the structural evaluation for the RT-100 under the RT Quality Assurance Program [Ref. 1] and summarizes the results to demonstrate compliance with the structural requirements of 10 CFR Part 71 [Ref. 2]. These evaluations follow nuclear industry standards

[Refs. 3 - 20]. Chapter 1 General Information and Chapter 3 Thermal Evaluation provide input to the Chapter 2 Structural Evaluation; furthermore, these three chapters feed information to later Chapters of the SAR as demonstrated in Figure 2-1 on the following page.

The RT-100 structural performance under 10 CFR Part 71 [Ref. 2] Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) significantly affects the package ability to meet the thermal, containment, shielding and subcriticality requirements. Consequently, results from the structural evaluation are used in the thermal, containment, and shielding evaluations (Note: criticality issues are not applicable to the RT-100).

The foremost structural requirement of the RT-100 is to withstand NCT and HAC loadings with sufficient structural integrity to maintain shielded containment. Evaluations in the following sections demonstrate the RT-100 package design satisfies these requirements. Before presenting these detailed evaluations, a general description of the RT-100 cask design is provided and includes complete specifications for the containment boundary.

2.1 Description of Structural Design Major design features that govern the structural performance of the RT-100 under NCT and HAC conditions are the impact limiters (upper and lower) and the cask body including the impact limiter attachment rings, bolting ring, primary and secondary lids, lifting pockets and tie- down arms.

These features are sufficiently designed so that the structural response of the RT-100 exceeds all 10 CFR 71 [Ref. 2] requirements.

Appendix 1.4 (Attachment 1.4-2 thru 1.4-8) shows the general assembly drawings of the RT-100 Cask Package. The major components are identified and include the impact limiters and cask body.

As subsequently discussed in Section 2.1.1.1, the package containment boundary is defined by the inner surfaces of the cask body, and the primary and secondary lids. Shielding is provided by the following features:

o Cask bottom and sidewall that contain 75 and 90 mm lead layers, respectively o 210 mm thick stainless steel primary lid o 170 mm (nominally) stainless steel secondary lid with embedded 60 mm thick lead layer Robatel Technologies, LLC Page 2-1

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2-1 Information Flow for the Structural Review Structural Review Loading

Compression

Crush

Impact

Load Combinations

Penetration

Pressure

Puncture

Vibration

Water Spray

Thermal Evaluation

Compression

Stress Analysis

Physical Testing Results

Buckling

Fatigue

Strain and

Stress

Fracture

Water In-Leakage Deformation Thermal Containment Shielding Criticality Evaluation Evaluation Evaluation Evaluation

Deformation

Deformation of the

Package

Not Applicable

Crushing/Puncture Containment Deformation Boundary

Crushing/Puncture

Chemical and

Extrusion Galvanic Reactions

Slump

Contents Condition

Displacement of Contents and Shielding Operating Acceptance Tests Procedures and Maintenance

Closure

Codes and Requirements Standards

Bolt Torque

Pressure and

Loading Structural Tests Configuration

Component Tests

Tie-Down Configuration

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.1.1 Discussion The RT-100 cask body is a cylindrical container with an outside diameter of 2060 mm and an overall height of 2321 mm (including lids). The sidewalls are nominally 165 mm thick, consist of a 90 mm thick lead layer encased by 30 mm thick internal and 35 mm thick external (ASTM A240, Type 304) stainless steel shells, have a 5 mm thick ceramic insulation layer, and have an outer 5 mm thick protective shell (ASTM A240, Type 304L stainless steel). The cask sidewall design varies from the above description in the following areas:

o Regions of the cask body encompassed by the impact limiters o Impact limiter attachment rings o Lifting pocket locations o Tie-down arm attachment pads.

The specific sidewall configuration at each of these locations is further described and fully considered in all subsequent evaluations.

The bottom end of the cask body consists of a 75 mm thick lead layer encased by a 50 mm thick (ASTM A240, Type 304L) stainless steel bottom forging on top, and a 30 mm thick external stainless steel bottom plate underneath. The bottom forging is connected to the inner shell with full penetration welds. The bottom plate is connected to the outer shell with a full penetration weld.

The top end of the cask body consists of an upper forging (ASTM A240, Type 304L), and two lids (primary and secondary, both ASTM A240, Type304L). The upper forging is connected to the inner shell with full penetration welds. The upper forging is connected to the cask outer shell with full penetration welds. Thirty-two (32) M48x2d threaded holes for securing the primary lid are equally spaced along the upper forging top surface. The upper forging top surface also provides a seating surface for the primary lid seals. The primary lid is nominally 210 mm thick.

The primary lid has thirty-two (32) clearance holes near its outer periphery for the M48 bolts (ASTM A354 Gr. BD or equivalent), which secure it to the bolting ring. These clearance holes are sufficiently counter-bored to preclude direct impact to the M48 bolts during a drop.

Additionally, the primary lid has a central 737 mm diameter through-hole with a 2016 mm OD x 82 mm deep counter-bore. The counter-bore surface has eighteen (18) M36x2d equally spaced threaded holes for securing the secondary lid and also provides a seating surface for the secondary lid seals. The secondary lid is nominally 170 mm thick with an embedded 60 mm thick lead layer.

The secondary lid has eighteen (18) clearance holes near its outer periphery for the M36 bolts (ASTM A354 Gr. BD or equivalent) used to attach it to the primary lid. The primary and secondary lids have one vent port each which allows for leakage monitoring.

The impact limiters are cylindrically-shaped components that surround the top and bottom ends of the cask as shown in Chapter 1, Figure 1.2.1-1. Each impact limiter has twelve (12) M36 studs.

The impact limiters are attached to the cask with these studs that pass through clearance holes in the top and bottom impact limiter attachment rings, and accept M36 stainless steel nuts. The impact limiters are comprised of segmented polyurethane foam blocks encased in relatively thin stainless Robatel Technologies, LLC Page 2-3

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 steel outer coverings. The outer coverings are 4 mm thick except near the cask surface where the thickness is 10 mm. During NCT and HAC tests, the impact limiters are designed to protect the cask by absorbing energy and for providing thermal insulation.

2.1.1.1 Containment Boundary As shown in Chapter 4, Figure 4.1.2-1 (Illustration of Containment Boundary), the containment boundary of the RT-100 cask is defined by the following specific features of the cask body and the primary and secondary lid.

o Bottom forging at the bottom end of the cask o Inner shell that forms the wall of the cask with a full penetration weld o Full penetration weld between the inner bottom forging and the inner shell bottom o Top forging at the top of the cask o Full penetration weld between the upper forging and inner shell top o Primary lid and inner O-ring o Vent port cover plate and inner O-ring o Secondary lid and inner O-ring 2.1.2 Design Criteria The RT-100 design satisfies the NCT requirements of 10 CFR 71.71 [Ref. 2], and HAC requirements of 10 CFR 71.73 [Ref. 2]. Furthermore, the design complies with General Standards for All Packages as specified in 10 CFR 71.43 [Ref. 2], and the Lifting and Tie- Down Standards specified in 10 CFR 71.45 [Ref. 2]. These criteria are demonstrated in Sections 2.5.1 and 2.5.2.

The design criteria used in the qualification of the RT-100 were selected based on guidance provided in Regulatory Guide 7.6 [Ref. 4]. Regulatory Guide 7.6 provides design criteria based on the ASME B&PV Code,Section III [Ref.7], and is intended for Type B packages used to transport irradiated fuel assemblies. Therefore, allowable stresses values for NCT Service Level A Limits and HAC Service Level D Limits are conservatively adopted from Regulatory Guide 7.6

[Ref. 4] for the qualification of the RT-100 cask body.

Allowable stresses are derived from the Stress Intensity values appropriate to ASME B&PV Code,Section III, Subsection ND [Ref. 7]. Stress Intensity values based on Subsection ND are presented in Table 2.2.1-1.

The load combinations used in performing the structural evaluations of the RT-100 cask are in accordance with Regulatory Guide 7.8 [Ref. 3]. Load combinations for the RT-100 cask body analysis are summarized in Table 2.1.2-1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.1.2-1 Load Combinations for RT-100 Cask Body Analyses LOAD NORMAL ACCIDENT Reg. Guide 7.8 A D Load Combinations 1 2 1 2 3 4 Dead Weight With maximum contents X X X X X X Thermal Hot X X X Stresses Cold X X X Normal X X X X Internal Pressure Accident (fire) X X Drop/Impact 0.3 Meters X X Drop/Impact 9 Meters X X 2.1.2.1 Cask Body Criteria (except Bolts and O-Rings)

The criteria for the cask shells and lids are developed per Regulatory Guide 7.6 Regulatory Position 2 [Ref. 4]. (The tie-down arms are also fabricated from stainless steel but their criteria are developed separately in Section 2.5.2). Table 2.1.2-2 provides a summary of the allowable stress limits defined in Regulatory Guide 7.6.

Table 2.1.2-2 Structural Design Criteria for RT-100 Reg. Guide 7.6 Service Level Stress Criteria Notes Pm Sm (1)(2)

Normal conditions: Service Level A Pm + Pb 1.5 Sm (2)

Pm + Pb + Q 3 Sm (3)

Pm 2.4 Sm or 0.7 Su (whichever is less) (4)

Accident conditions: Service Level D Pm + Pb 3.6 Sm or 1.0 Su (whichever is less) (4)

Total Stress < 2 Su (5)

1. Regulatory Guide 7.6 [Ref. 4], Regulatory Position 1

2. Regulatory Guide 7.6, Regulatory Position 2

3. Regulatory Guide 7.6, Regulatory Position 4

4. Regulatory Guide 7.6, Regulatory Position 6

5. Regulatory Guide 7.6, Regulatory Position 7 2.1.2.2 Bolts The allowable stresses under NCT (per NUREG/CR-6007 [Ref. 10]) are:

f t < Sm f t max < 3Sm if Su < 689 MPa

< 2.7Sm if Su > 689 MPa Pm + Pb + residual torsion < Sm where ft = average tensile stress Robatel Technologies, LLC Page 2-5

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 f t max = maximum tensile stress under combined tension and bending, and all other terms are as previously defined.

The allowable stresses under NCT (per NUREG/CR-6007 [Ref. 10]) are:

ft < Ftb fv < Fvb 2 2 ft fv

+ < 1.0 Ftb Fvb where fv = average shear stress Ftb = allowable average tensile stress

= Min (0.7Su, Sy) at temperature Fvb = allowable average shear stress

= Min (0.42Su, 0.6Sy) at temperature and all other terms are as previously defined.

2.1.2.3 Lead The structural integrity of the RT-100 cask does not depend on lead strength and thus, no lead strength criteria are specified. Mechanical and thermal properties which are important to the RT-100 cask structural performance are discussed in Sections 2.2, 2.14, and 3.2 2.1.2.4 Foam Criteria of the polyurethane foam used in the impact limiters are provided in Appendix 2.12 Impact Limiter Evaluation.

2.1.3 Weights and Centers of Gravity The nominal RT-100 weights and centers of gravity are shown in Table 2.1.3-1. Refer to RT100 PE 1001-1 Rev. H - Robatel Transport Package RT-100 General Assembly Sheet 1/2 (Chapter 1, Appendix 1.4, Attachment 1.4-2) for identification of assemblies and centers of gravity data. These weights are utilized in the structural evaluation presented in this chapter.

With the exception of the impact limiter, all analyses are performed with no less than a minimum gross weight of 41,500 kg. The impact limiter calculation is performed using 41,000 kg. The reason for this is that the max crush is obtained by using the minimum density of the foam. The calculation package RTL-001-CALC-ST-0401 Rev. 6 [Ref. 40] calculates the maximum g-load using both 41,500 kg and 41,000 kg. It is shown that max g-load is obtained using a gross weight of 41,000 kg. Thus, the impact limiter calculation is performed using a gross weight of 41,000 kg.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.1.3-1 Assembly Weights and Center of Gravity Locations Nominal Weight Center of Gravity3 Assembly3 (kg) (mm)

Lower Impact Limiter 2,450 516 Cask Body 24,500 1,446 Primary Lid w/ bolts 3,670 2,716 Secondary Lid w/ bolts 870 2,737 Upper Impact Limiter 2,550 2,812 Total Assembly Empty 34,040 1,650 1,434 min.3 Payload 6,8051 1,826 max.3 1,620 min.3 Total Assembly with payload 40,8452 1,676 max.3 Notes: 1. Maximum.

2. A minimum weight of 41,000 kg was used in all structural evaluations.

3. Value determined using payload center of gravity at 10% of cask interior height below or above the cask interior geometric centerline.

As shown in Table 2.1.3-1, the center of gravity of the empty RT-100 cask is approximately 1650 mm above the bottom of the cask. This location is just 20 mm lower than the 1630 mm elevation of the center of the inner cavity. Further, the maximum payload weight is less than 17%

(= 6,805/40,845 x100%) of the loaded cask weight. Thus, payload weight and/or center of gravity variations will not result in large changes to the loaded RT-100 cask center of gravity. Indeed, locating the payload center of gravity within 10% of the cavity internal height above or below the cavity centerline elevation moves the loaded RT-100 cask center of gravity by no more than +/- 28 mm. Such minor variations are insignificant during either NCT or HAC.

2.1.4 Identification of Codes and Standards for Package Design Since the package is used to transport contents with 3,000 A2 (as defined in 10 CFR 71.4 [Ref. 2]),

the RT-100 cask is a Type B Category II package per Regulatory Guide 7.11 [Ref. 5]. The codes and standards used in the design of the RT-100 cask are selected based on guidance provided in Regulatory Guide 7.6 [Ref. 4 and NUREG/CR-3854 [Ref. 6] for packages transporting Category II contents.

Per NUREG/CR-3854 [Ref. 6], the package containment system is fabricated in accordance with the ASME Code,Section III, Subsection ND [Ref. 7], and the tie-downs are fabricated in accordance with Subsection NF [Ref. 8]. These codes are applicable to the RT-100 cask design as they were developed for components of similar material as well as, for similar loading operations and potential package failures.

Several regulatory guides and NUREGs are used to design and evaluate the RT-100 package.

Regulatory Guide 7.8 [Ref. 3] is used in identifying the load combinations to be used in package design evaluation. Regulatory Guide 7.6 [Ref. 4] is used to determine the design criteria.

NUREG/CR-4554 [Ref. 9] is used in evaluating buckling of the containment vessel. NUREG/CR-Robatel Technologies, LLC Page 2-7

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 6007 [Ref. 10] is followed for the bolt evaluations.

2.2 Materials Material properties used in the RT-100 cask structural analyses are shown in Tables 2.2.1-1, 2.2.1-2, and 2.2.1-3. Material properties for the structural analyses of the polyurethane foam used in the impact limiter evaluations are provided in Appendix 2.12. Properties of both cask materials and foam used in the thermal analyses are provided in Section 3.2.1.

2.2.1 Material Properties and Specifications Structural components of the cask body are specified to be ASME A240 Type 304/304L steel, with the exception of the tie-down straps, which are ASME A240 UNS No. S31803 (Type 318) stainless steel. The primary and secondary lids are ASME A240 Type 304/304L steel, and the M36 and M48 bolts used to secure the lids are fabricated to meet the critical characteristics given in Chapter 8. These materials meet the requirements of ASME Section III, Subsection ND [Ref. 7].

Strength properties for these materials are presented in Table 2.2.1-1 using material information taken from ASME Section II-D [Ref. 31]. Table 2.2.1-2 provides density and Poissons ratio values also from ASME Section II-D.

The shielding is specified to be ASTM B-29 lead. The lead properties are provided in NUREG/CR-0481[Ref. 11] and are presented in Table 2.2.1-2.

EPDM (material designation per ASTM D1418) is used for all O-rings as part of the containment boundary. They serve as one of the boundaries for the cask. These O-rings have a usable temperature range going from -50°C up to 150°C; this temperature range meets or exceeds both NCT and HAC requirements.

RT verifies that all the materials of structural components have sufficient fracture toughness to preclude brittle fracture under NCT and HAC. Regulatory Guides 7.11 [Ref. 5] and 7.12 [Ref. 16]

are used to provide criteria for fracture toughness. RT shall procure all materials under the RT Quality Assurance Program [Ref. 1] with the specifications for each material. Regulatory Guides 7.11 and 7.12 do not apply to the RT-100; use of Stainless Steel ASTM A-240 type 304, ASTM A-240 type 304L, and ASTM A-240 UNS S31803 precludes brittle fracture under both NCT and HAC.

RT verifies that all material properties are appropriate for the load conditions specified in Regulatory Guide 7.6 [Ref. 4] and temperatures at which allowable stress limits are defined are consistent with minimum and maximum service temperatures. Allowable stresses based on Regulatory Guide 7.6 [Ref. 4] at the bounding NCT temperature of 100°C are provided in Table 2.2.1-3. Allowable stress intensities at other temperatures considered to be the bounding condition for a specific case are defined as needed in the section where that analysis is presented.

RT verifies that all the force-deformation properties for impact limiters are based on appropriate test conditions and temperature. Test parameters for qualifying the foam material are identified in Chapter 2, Appendix 2.13.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.2.1-1 Cask Temperature-Dependent Material Properties Yield Tensile Design Stress Coefficient of Strength Strength Intensity Young's Thermal Material Temperature (Sy) (Su) (Sm) Modulus Expansion

(°C) (MPa) (GPa) (10-6 /°C)

-30 207 517 138 198 20 207 517 138 195 15.3 ASME SA-240 65 184 496 138 192 15.8 Type 304/304L 100 170 485 138 189 16.2 (Dual Certified) 150 154 456 138 186 16.6 200 144 442 129 183 17.0 250 135 437 122 179 17.4

-30 172 483 115 198 20 172 483 115 195 15.3 ASME SA-240 65 157 463 115 192 15.8 Type 304L 100 146 452 115 189 16.2 150 132 421 115 186 16.6 200 121 406 110 183 17.0 250 114 398 103 179 17.4

-30 172 483 115 198 20 172 483 115 195 15.3 65 157 471 106 192 15.8 ASME SA-240 100 145 467 96.3 189 16.2 Type 316L 150 131 441 87.4 186 16.6 200 121 429 81.2 183 17.0 250 114 426 76.0 179 17.4

-30 448 621 207 = Su/3 211 20 448 621 207 205 15.3 65 418 620 207 200 15.8 ASME SA-240 100 395 619 206 194 16.2 UNS No. S31803 150 370 598 199 190 16.6 200 354 577 193 186 17.0 250 344 564 188 183 17.4

-30 896 1030 343 = Su/3 199 20 896 1030 343 202 11.5 ASME SA-354 65 855 1030 343 199 11.8 Grade BD (Bolting 100 816 1030 343 197 12.1 material) 150 792 1030 343 194 12.4 200 768 1030 343 191 12.7 250 737 1030 343 188 13.0 ASME SA-479, ER308 -30 to 40 205 515

-29 16.75 28.2 20 15.67 28.9 50 14.94 29.4 ASTM B-29 Lead 100 13.73 30.2 150 12.74 31.2 200 11.80 32.6 250 10.70 34.1 Robatel Technologies, LLC Page 2-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.2.1-2 Cask Temperature-Independent Material Properties ASME [Ref. 31]

Density Poissons Material (kg/m³) Ratio ASME SA-240 Type 304/304L (Dual Certified) 8030 0.31 ASME SA-240 UNS No. S31803 8030 0.31 ASME SA-354 Grade BD (Bolting material) 7750 0.30 ASTM B-29 Lead 11300 0.40 Table 2.2.1-3 Allowable Stresses for Cask Body Materials Material ASME SA-240 ASME SA-240 ASME SA-240 ASME SA-240 ASME SA-354 Design Criteria Type 304/304L UNS No.

Type 304L Type 316L Grade BD (Dual Certified) S31803 MPa MPa MPa MPa MPa Yield Stress, Sy 170 146 145 395 816 Tensile Strength, Su 485 452 467 619 1030 Design Stress Intensity, Sm 138 115 96.3 206 299 Pm 138 115 96.3 206 299 Normal Pm + Pb 207 173 144 309 449 Conditions Pm + Pb + Q 414 345 289 618 897 Hypothetical Pm 331 276 231 433 718 Accident Pm + Pb 485 414 347 619 1030 Conditions Total Stress 970 904 934 1238 2060 2.2.2 Chemical, Galvanic, or Other Reactions The materials used in the fabrication and operation of the RT-100, including coatings, lubricants, and cleaning agents, are evaluated to determine whether chemical, galvanic, or other reactions among the materials, contents, and environments can occur. All phases of operation, loading, unloading, handling, storage, and transportation, are considered (in conjunction with the procedures described in Chapter 7) for the environments that may be encountered under normal, off-normal, or accident conditions. Based on the evaluation, there are no potential reactions that could adversely affect the overall integrity of the cask or the structural integrity and retrievability of the contents from the cask. The evaluation conforms to the guidelines of NRC Bulletin 96-04, Chemical, Galvanic, or Other Reactions in spent Fuel Storage and Transportation Casks, dated July 5, 1996 [Ref. 52], and demonstrates that the RT-100 cask meets the requirements of 10 CFR 71.43(d) [Ref. 2].

2.2.2.1 Component Material Categories The component materials evaluated are categorized based on similarity of physical and chemical properties and/or on similarity of component functions. The categories of materials that are considered are as follows:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Stainless/nickel alloy steels o Nonferrous metals o Shielding materials o Criticality control materials o Energy absorbing materials o Cellular foams and insulations o Lubricants and greases o O-rings o Secondary Containers and Shoring o Filters These categories are evaluated based on the environment to which they could be exposed during operation or use of the RT-100.

The RT-100 component materials are not reactive among themselves, with the casks contents, nor with the casks operating environments during any phase of normal, or accident condition loading, unloading, handling, storage or transportation operations. No reactions occur, and no gases or other corrosion byproducts are generated.

2.2.2.1.1 Stainless/Nickel Alloy Steels No reaction of the cask components (stainless or nickel alloy) is expected in any environment.

During the fabrication process of the RT-100 ridges and crevices on the external surfaces are reduced through the finishing process and the external surface is passivated to prevent corrosion.

Galvanic corrosion between the stainless steels and nickel alloy steels does not occur due to the lack of effective electrochemical potential difference between these metals. No coatings are applied to the stainless steel or nickel alloy steels.

There is no potential for a reaction between stainless steel and any silicone products, fluorocarbon elastomers, dry film lubricants, blended polytetrafluoroethylene (PTFE), or ethylene glycol.

Based on the foregoing discussion, there are no potential reactions expected with the stainless steel cask components.

2.2.2.1.2 Nonferrous Metals There are no nonferrous metals used in the RT-100. Therefore, no electrochemical driving potential exists.

2.2.2.1.3 Shielding Materials The primary shielding materials used in the RT-100 is lead which is completely enclosed and sealed in stainless steel. Therefore, there are no potential reactions associated with the cask shielding materials.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.2.2.1.4 Criticality Control Material The RT-100 does not contain materials for criticality control. Therefore, no potential reactions associated with these materials exist.

2.2.2.1.5 Energy Absorbing Material The RT-100 utilizes polyurethane foam for energy absorption in the impact limiters. The foam is completely enclosed (sealed) in stainless steel and there are no potential reactions between the foam and the stainless steel shells. The foam is cured, cut, and machined prior to installation. During fabrication the machined foam blocks are inserted into the impact limiter stainless steel shell.

During the welding process backing strips, high temperature heat tape, and rock wool are used to protect the foam. Therefore, no potential reactions associated with the energy absorbing material exists.

2.2.2.1.6 Cellular Foam and Insulation The RT-100 does not utilize cellular foam or insulation. Therefore, no potential reactions associated with the cellular foam or insulation exists.

2.2.2.1.7 Lubricant and Grease The dry film lubricants used with the RT-100 meet the performance and general compositional requirements of the nuclear power industry. These lubricants are used primarily on threaded/mechanical connection surfaces. These lubricants are insoluble in most solutions. There are no potential reactions associated with these lubricants or grease.

2.2.2.1.8 O-Rings The RT-100 utilizes seals formed from EPDM. EPDM is a synthetic rubber elastomer. Elastomer O-rings are used for transport cask applications because of their excellent short-term sealing capabilities, ease of handling, and more economical cost. Seal and gasket materials have stable, non-reactive compositions. There are no potential reactions associated with the RT-100 seal materials.

2.2.2.1.9 Secondary Containers and Shoring Secondary containers and shoring features may be constructed of carbon steel, stainless steel, wood, or a thermoplastic such as polyethylene or polypropylene.

2.2.2.1.10 Filters Filters shipped for disposal may be constructed from stainless steel or thermoplastic such as polyethylene or polypropylene.

2.2.2.2 General Effects of Identified Reactions No significant potential galvanic or other reactions have been identified for the RT-100. Therefore, no adverse conditions can result during any phase of cask operations for NCT or HAC.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.2.2.3 Adequacy of the Cask Operating Procedures Based on the results of this evaluation, it is concluded that the RT-100 operating controls and procedures presented in Chapter 7 are adequate to minimize occurrence of hazardous conditions.

2.2.2.4 Effects of Reaction Products No significant potential chemical, galvanic, or other reactions are identified for the RT-100.

Therefore, the overall integrity of the cask and the structural integrity and retrievability of the contents are not adversely affected for any cask operations throughout the design basis life of the cask. Based on the evaluation, no significant reactions are identified and thus, there is no change in cask properties, no binding of mechanical surface, and no degradation of any safety components either directly or indirectly.

2.2.3 Effects of Radiation on Materials Gamma radiation has no significant effect on metal and therefore, the radiation produced by the contained radioactivity does not cause any measurable damage to the cask metallic components (stainless steel, carbon steel and lead).

For seals, the absorbed dose in a year is expected to be below 350 rad which is significantly below the polymer damage threshold of 1x105 rad. Additional support information about EPDM resistance to radiation up to 5x108 rads while retaining reasonable flexibility and strength, hardness and very good compression set resistance is provided by an IEEE paper [Ref. 54].

For the ceramic thermal shield, the absorbed dose is expected to be below 350 rad. However, ceramic materials are insensitive to gamma radiation damage and thus, the ceramic thermal shield is expected to be unaffected by radiation.

2.3 Fabrication and Examination The following subsections provide a summary description of fabrication and examination of the RT-100. A more detailed description is provided in subsequent sections of the SAR.

2.3.1 Fabrication The RT-100 packaging is designed as a category II container, as mentioned in Section 2.1.4.

Fabrication and procurement of the containment components is based on ASME B&PV code, section III, Subsection ND [Ref. 7]. The other components (non-containment) are fabricated based on ASME B&PV code,Section III, subsection NF [Ref.8]. See Sections 2.1.2 and 2.1.4 for additional information.

2.3.2 Examination Examination of the RT-100 during and after fabrication is conducted in accordance with the requirements of the ASME B&PV code,Section III, Subsection ND-5000 [Ref. 7]. The non-containment components examination is conducted in accordance with the requirements of ASME B&PV Code,Section III, Subsection ND-5000 or NF5000 [Ref. 8]. See Chapter 8, Sections 8.1 and 8.2 for additional information.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.4 General Requirements for All Packages The RT-100 meets or exceeds all the requirements in 10 CFR 71.43 [Ref 2]. Also, the RT-100 meets the general package requirements Regulatory Guide 7.9 [Ref. 49] as listed below:

o Smallest overall dimension is greater than 10 cm (4 in).

o Outside of the cask incorporates a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons.

o Cask includes a containment system closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package.

The following sections describe compliance of the RT-100 with these requirements.

2.4.1 Minimum Package Size This section is not applicable since the RT-100 has dimensions larger than 10 cm (4 inches). The smallest overall dimension of the cask body is the outer diameter, which is over 200 cm.

2.4.2 Tamper-Indicating Feature The RT-100 upper impact limiter covers the upper end of the cask including the primary and secondary lids, which prevents access to the cask lids. Therefore, tamper-indicating devices are attached to the impact limiter aligning pin. Impact limiters are installed on the cask body following the lid closure operation. Once the impact limiters are installed on the cask body, the attachment nuts are threaded on the attaching studs and hand-tightened (drop testing has shown that torquing of the attachment bolts is not necessary). A tamper-indicating seal is installed on the aligning pin of the upper impact limiter to ensure that removal of the impact limiter by unauthorized individuals can be detected.

2.4.3 Positive Closure The RT-100 design includes a containment system that is bounded by the inner shell, primary lid, secondary lid, and vent port cover plate. Each lid and the cover plate are secured to the cask body by multiple bolts. These bolts are tightened during the loading process to a set torque value that cannot be inadvertently loosened. Additionally, the stress analysis of the bolts presented in Section 2.6.7 demonstrates that the bolts can maintain positive closure during operation.

2.5 Lifting and Tie-Down Standards for All Packages The RT-100 lifting and tie-down components are evaluated structurally in the following sections.

The lifting and tie-down requirements are as specified in 10 CFR 71.45 [Ref. 2].

2.5.1 Lifting Devices The primary lifting device for the RT-100 is the set of two lifting pockets that are welded to the outer shell of the cask. After removal of the impact limiters, the lifting pockets are designed to allow the loaded cask to be lifted using a lifting yoke. The primary and secondary lids and the upper/lower impact limiters are fitted with threaded bolt holes; these bolt holes provide for attachment of lifting rings that are used in lifting each component.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.5.1.1 Lifting Design Criteria Lifting attachments that are a structural part of the RT-100 cask are designed with a minimum safety factor of three against yielding when used to lift the package. The lifting devices are also designed so that any failure of the lifting device under excessive load would not impair the ability of the RT-100 to meet other requirements of 10 CFR 71.45 [Ref. 2]. The design weights used in the lifting evaluation are as follows:

o Fully loaded RT-100 with maximum contents and the lower impact limiter is 41,500 kg o Primary lid with secondary lid in place is 4,505 kg o Secondary lid is 857 kg o Upper impact limiter is 2,541 kg o Lower impact limiter is 2,448 kg 2.5.1.2 Lifting Device Descriptions In this section, the following RT-100 components are evaluated for lifting:

o Lifting Pockets o Primary Lid o Secondary Lid o Lower Impact Limiter o Upper Impact Limiter The lifting pockets are utilized to lift the assembled cask; the bounding configuration is the cask loaded with the maximum payload weight and the lower impact limiter attached. Additionally, the primary and secondary lids and the upper and lower impact limiters are evaluated for lifts using removable lifting rings.

2.5.1.3 Lifting Device Evaluations In the following sections, each device used for lifting is evaluated for stress. The details of each evaluation are presented including the worst-case stress results and safety factors. Additional details supporting these calculations are provided in Calculation Package RTL-001-CALC-ST-0201, Rev. 5 [Ref. 33].

2.5.1.3.1 Cask Body Lifting Evaluation The cask is lifted by using the two lifting pockets that are welded to the cask exterior sidewall on opposite sides of the cask body. The assembled and loaded cask is lifted with the upper impact limiter removed to accommodate the connection between the lift yoke and the lifting pockets. The cask lifting load is the total weight of the fully assembled cask, including the payload, but with the upper impact limiter load removed. The upper impact limiter is lifted separately. The lifting pockets are evaluated for the tear-out stress, bearing stress, and weld stress due to the required lifting activities. The lifting pockets are also evaluated for pure shear stress as described in ASME Section III Subsection NF [Ref. 8].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 A Dynamic Load Factor (DLF) of 1.35 is applied to the lift forces that act on the cask components during movement. ANSI N14.6 [Ref. 56] requires additional safety features for handling of critical loads. One option identified is to apply increased stress design factors on the load-bearing members; however, the standard does not recommend a value for the stress design factor. The German Nuclear Safety Standards Commission provides standard KTA-3905 for lifting loads in nuclear power plants. [Ref. 57] This standard requires a live load factor of 1.35 for dead weight lifts. This calculation uses the KTA-3905 live load factor value as the dynamic load factor. The dynamic load factor is applied to all load bearing members.

2.5.1.3.1.1 Lifting Pocket Design Features The lifting pockets are manufactured from blocks of ASTM A240 Dual Certified Type 304/304L stainless steel that are welded to opposite sides of the outer shell of the cask body, also manufactured from ASTM A240 Type 304/304L stainless steel. The weld material is SA-279 Grade ER308 UNS S30880. The welds extend down both sides and along the bottom of the lifting pockets, forming a U shape. The lifting pockets have a cutout that allows the lifting yoke to pass downward and through the lifting pocket. The connection is completed with a rectangular shaped retaining pin that is inserted through cutouts in both the lifting pocket and the lifting yoke.

Figure 2.5.1-1 provides the configuration and dimensions of the lifting pockets and shows the cutouts for the lifting yoke and retaining pin. The design loads and material strengths of the lifting pocket base metal and weld materials are as follows:

Total Lifted Cask Weight W= 41,500 - 2,541 kg = 38,959 use 39,500 kg Dynamic Load Factor DLF = 1.35 Number of Lifting Pockets np = 2 Gravitational Acceleration g = 9.81 m/s² WxDLFxg 39500x1.35x9.81 1 kN Vertical Shear Load PV = = x np 2 1000 N

= 261.6 kN pocket Lifting Pocket Yield Strength Sy = 199 MPa Lifting Pocket Tensile Strength Su = 511 MPa Factor of Safety on Yield Strength Fsy = 3 Factor of Safety on Tensile Strength Fsu = 5 The critical dimensions for the weld evaluation are as follows. These dimensions ignore the dimensions of the welds.

Lifting Pocket Length Lp = 191 mm = 0.191 m Lifting Pocket Edge Distance dp = 55 mm = 0.055 m Lifting Pocket Eye Length Le = 84 mm = 0.084 m Retaining Pin Dimensions Wp= 60 mm = 0.060 m Hp = 80 mm = 0.080 m Robatel Technologies, LLC Page 2-16

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The eye refers to the rectangular cutout in the lifting pocket for the retaining pin and the eye length is the vertical height of the eye. The lifting pocket length is the distance from the horizontal centerline of the retaining pin eye to the top of the lifting pocket. The lifting pocket edge distance refers to the vertical height of the recessed cap on the lifting pocket.

Figure 2.5.1-1 RT-100 Lifting Pocket Dimensions 2.5.1.3.1.2 Lifting Pocket Tear-out Stresses The lifting pockets are used for lifting the assembled and loaded cask body, without the upper impact limiter, and are rendered inoperable by removing the lifting attachment from the lifting pocket during transport. The lifting pockets are considered to be a structural part of the package with respect to lifting and shall be designed for the factor of safety against yielding and ultimate stresses. A lifting yoke is used to lift the assembled cask body and to ensure that the lifting straps or cables remain parallel to the body of the cask during lifting operations. The tear-out stresses for the lifting pocket retaining pin hole are as follows:

0.084 Lifting Eye Tear-out distance = = 0.191 0.055 2 2

= 0.094 m Lifting Pocket Thickness tp = 110.5 - 40 = 70.5 mm = 0.071 m Lifting eye Tear-out Area = x = 0.094 0.071

= 0.00663 m² The tear-out stresses for the lifting pocket are calculated:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 261.6 Nominal Tear-out Stress = = = 19734 = 19.7 2 x 2 x0.00663 2 Allowable Yield Stress y = 0.6 x SyL = 119 MPa Allowable Ultimate Stress u = 0.6 x SuL = 307MPa Factor of Safety on Yield Strength 119

= = = 6.05 > 3.0 19.7 Factor of Safety on Tensile Strength 307

= = = 15.54 > 5.0 19.7 2.5.1.3.1.3 Lifting Pocket Bearing Stresses The bearing stress in the lifting pocket from the lift yoke retaining pin is calculated as follows.

The acceptance criterion for the pocket bearing stress are the yield strength of the material.

Lifting Pocket Bearing Area

= x = 0.06 x 0.071 = 0.00423 2 Nominal Bearing Stress 261.6

=

= 0.00423

= 61834 2

= 61.8 Factor of Safety on Yield Strength 199

= = = 3.22 > 1.0 61.8 2.5.1.3.1.4 Lifting Pocket Weld Stresses The stresses in the welds (attaching the lifting pocket to the cask outer shell) are found by applying the shear load from the lifting pockets to the weld around the perimeter of the plate. Based on the safety factors for the lifting pocket, yielding controls the weld evaluation. The stresses and allowables are determined as described in Design of Welded Structures [Ref. 25] and Calculation Package RTL-001-CALC-ST-0201, Rev. 5 [Ref. 33]

Conservatively, the upper section of the pocket is considered to take the full lifting load. The lifting pocket is seal welded to and bears upon the cask bolting ring. The lifting load is therefore shared between the lifting pocket weld and the bolting ring. Conservatively, the full load is considered to be taken by the lifting pocket weld only.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The stresses in the welds attaching the lifting pocket to the cask outer shell are found by applying the shear load from the lifting pockets to the weld around the perimeter of the lifting pocket. Based on the safety factors for the lifting pocket, yielding controls the weld evaluation. The welds on the lifting pockets are evaluated as a line force on the weld as described in Design of Welded Structures [Ref. 25] (Refer to pages 7.4-6 and 7, Tables 4 and 5). Since the cask is lifted using a yoke that maintains the force in a vertical direction, there are no bending or twisting loads, so the section Modulus and the polar moment of inertia are zero and can be ignored. The weld geometry is provided in Figure 2.5.1-2 Y

Cx X X Base Metal Thickness d

Weld Throat Size Y

Cy Y

X b Local Coordinates Figure 2.5.1-2 Weld Geometry Weld properties are as follows:

Length of horizontal weld b = 0.28 m Length of vertical weld d = 0.20 m Weld Length Aw= b + 2d = 0.68 m Weld Throat Size Tw= 0.015 m Base Metal (Cask Wall) Thickness Tc= 0.035 m The force acting on the weld is:

Fy 261.6 kN fvy = Aw

= 0.68

= 384.71 m Yield Weld Allowable wya = 0.6 x Swy x Tw x 1000

= 0.6 x 205 x 0.015 x 1000 = 1845 kN/m Tensile Weld Allowable wya = 0.6 x Swu x Tw x 1000

0.6 x 515 x 0.015 x 1000 = 4635 kN/m Yield Cask Allowable 0.6 x Scy x Tc x 1000 0.6 x 199 x 0.035 x 1000 cya =

0.7071 0.7071

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Tensile Cask Allowable 0.6 x Sc x Tc x 1000 0.6 x 511 x 0.036 x 1000 cua = =

0.7071 0.7071

= 15176 kN/m Weld Yield FS wya 1845

= = = 4.80 > 3.0 fw 384.71 Weld Tensile FS wua 4635

= = = 12.05 > 5.0 fw 384.71 Cask Yield FS cya 5910

= = = 15.36 > 3.0 fw 384.71 Cask Ultimate FS cua 15176

= = = 39.45 > 5.0 fw 384.71 2.5.1.3.1.5 Lifting Pocket Average Pure Shear The lifting pocket average pure shear is evaluated in accordance with ASME Section III Subsection NF [Ref. 8] Subparagraph 3223.2 and is limited to 0.6 Sm. The factor of safety is determined by comparing the pure shear to the lifting pocket tear out stress. For the lifting pocket weld evaluation, the average pure shear is evaluated as follows.

Cask Membrane Strength Sm = 115MPa Cask Allowable Pure Shear Sap = 0.6 x Sm = 0.6 x 115 = 69.0MPa FS for Cask Pure Shear 69.0

= = 19.7 = 3.50 > 1.0 2.5.1.3.1.6 Summary of Results Table 2.5.1-1 provides a summary of the Factors of Safety for each of the lifting conditions that are evaluated for the assembled RT-100. The table shows that all of the lifting conditions meet the required factor of safety greater than 3.0 against yield and the factor of safety greater than 5.0 against ultimate stress for the tear out and weld stress and a greater than 1.0 for the bearing stresses and average pure shear.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.5.1-1 Summary of Results for Lifting Assembled Cask Factor of Safety Lifting Condition Evaluated Yield Ultimate

(> 3) (>5)

Lifting Pocket Tear-out Stresses 6.05 15.54 Lifting Pocket Weld Stresses:

4.80 12.05 Weld Lifting Pocket Weld Stresses:

15.36 39.45 Cask Factor of Safety

(>1)

Lifting Pocket Bearing Stresses 3.22 N/A Lifting Pocket Average Pure 3.50 Shear 2.5.1.3.2 Primary Lid Lifting Evaluation The primary lid is evaluated for the working load limit in the lifting rings and for the tear-out stresses in the lid from the lifting activities. The lifting rings for the primary lid can only be used when the cask lid is separated from the cask body. The secondary cask lid is also removable, so the primary lid may be lifted with the secondary lid attached or separated from the primary lid.

Conservatively, the combined primary and secondary lid is used for the lifting evaluation. The primary lid design information is:

Primary Lid Weight WPL = 3648 kg, assume 3700 kg Secondary Lid Weight WSL = 857 kg, assume 900 kg Total Lid Lifting Weight WL = 3700 + 900 = 4600 kg Number of Lifting Rings nr = 3 Dynamic Load Factor DLF = 1.35 2.5.1.3.2.1 Primary Lid Lifting Ring Working Loads The lifting rings on the primary lid are only used for lifting when the lid is detached from the cask body, and are rendered inoperable by removing the rings from the lid when the cask is assembled.

The rings are therefore not considered to be a structural part of the package and do not need to be designed for the factor of safety against yielding.

WL xDLF 4600x1.35 Lifting Ring Load Pr = nr

= 3

= 2070 kg Ring Working Load Limit Pr,max = 3000 kg Pr,max 3000 Factor of Safety FS = Pr

= 2070

= 1.45 > 1.0 Robatel Technologies, LLC Page 2-21

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.5.1.3.2.2 Primary Lid Thread Engagement The minimum required thread engagement length is determined in accordance with Machinerys Handbook [Ref. 27]. The primary lid is manufactured from ASTM A240 Type 304L SS material.

This material is weaker than the M20 lifting ring material (ASTM A-354 Gr. BD), so failure will occur at the root of the primary lid material threads. The minimum required thread engagement length that prevents primary lid material failure is:

Minimum Engagement Length Le= Sbt 2 A b 1

Sct n Ds,min + 0.57735 ( Ds,min E n,max )

2 n Where Sbt = Bolt External Thread Tensile Strength, MPa Ab = Stress Area of Bolt External Threads, mm² Sct = Cask Internal Thread Tensile Strength, MPa n = Number of threads per millimeter Ds,min = Minimum Major Bolt Diameter, mm En,max = Maximum Pitch Diameter of Internal Thread, mm Solving the equation for Minimum Engagement Length, Le:

Minimum Engagement Length 150,000 x 2 x 0.38 Le =

1 69,000 x x 10.16 x 0.773 x [2 x 10.16 + 0.57735 x (0.773 0.699)]

= 0.73 in = 18.5 mm Where Sbt = 1030 MPa = 150,000 psi Ab = 245.0 mm² = 0.38 in² SLt = 470 MPa = 69,000 psi p = Thread Pitch = 2.5 mm = 0.098 in 1 1 n = = = 10.16 Threads/inch p 0.098 Ds,min = 19.623 mm = 0.773 in En,max = 17.744 mm = 0.699 in The available thread engagement, Lep, is 32 mm. Therefore, the factor of safety is:

Lep 32.0 FS = = = 1.73 > 1.0 Le 18.5 The lifting ring configuration is acceptable for the applied loads. In the unlikely event that failure does occur in the lid threads, no adverse effects on the RT-100 will occur since the threads are outside the cask containment boundary.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.5.1.3.3 Secondary Lid Lifting Evaluation The secondary lid is lifted using a set of three lifting rings that attach to threaded holes in the top surface of the lid. Although the maximum evaluated weight of the secondary lid lift includes only the secondary lid, the hardware is the same as that used for the primary lid. The combined primary and secondary lid are evaluated for lifting in Section 2.5.1.3.2. This section evaluates the working load limit in the lifting rings and for the minimum thread engagement in the lid during lifting activities. The secondary lid design information is:

Secondary Lid Weight WSL = 857 kg, assume 900 kg Number of Lifting Rings nr = 3 Dynamic Load Factor DLF = 1.35 2.5.1.3.3.1 Lifting Ring Working Load The lifting rings on the secondary lid are only used for lifting when the lid is detached from the cask and are rendered inoperable by removing the rings from the lid when the cask is assembled.

The rings are therefore not considered to be a structural part of the package and do not need to be designed for the factor of safety against yielding.

Lifting Ring Load WSLxDLF 900 x 1.35 Pr = = = 405 kg nr 3 Ring Working Load Limit Pr,max = 3000 kg Factor of Safety Pr,max 3000 FS = = = 7.4 > 1.0 Pr 405 2.5.1.3.3.2 Secondary Lid Thread Engagement The minimum required thread engagement length is determined in accordance with Machinerys Handbook [Ref. 27]. The secondary lid is manufactured from ASTM A240 Type 304L SS material. This material is weaker than the M20 lifting ring material (ASTM A-354 Gr. BD), so failure will occur at the root of the secondary lid material threads. The minimum required thread engagement length that prevents secondary lid material failure is:

Minimum Engagement Length Le = Sbt 2 A b 1

Sct n Ds,min + 0.57735 ( Ds,min E n,max )

2 n Sbt = Bolt External Thread Tensile Strength, MPa Ab = Stress Area of Bolt External Threads, mm² Sct = Cask Internal Thread Tensile Strength, MPa n = Number of threads per millimeter Ds,min = Minimum Major Bolt Diameter, mm En,max = Maximum Pitch Diameter of Internal Thread, mm Robatel Technologies, LLC Page 2-23

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Solving the equation for Minimum Engagement Length, Le:

Minimum Engagement Length 150,000 x 2 x 0.38 Le =

1 69,000 x x 10.16 x 0.773 x [2 x 10.16 + 0.57735 x (0.773 0.699)]

= 0.73 in = 18.5 mm Where Sbt = 1030 MPa = 150,000 psi Ab = 245.0 mm² = 0.38 in² SLt = 470 MPa = 69,000 psi p = Thread Pitch = 2.5 mm = 0.098 in 1 1 n = = = 10.16 Threads/inch p 0.098 Ds,min = 19.623 mm = 0.773 in En,max = 17.744 mm = 0.699 in The available thread engagement, Lep, is 32 mm. Therefore, the factor of safety is:

Lep 32.0 FS = = 18.5 = 1.73 > 1.0 Le Therefore, the secondary lid lifting ring configuration is acceptable for the required loads.

2.5.1.3.4 Upper Impact Limiter Lifting Evaluation The upper impact limiter is lifted using a set of three lifting rings that attach to threaded holes in the top surface of the limiter. The lifting rings are designed to remove the impact limiter from the cask body and not to lift the cask body while still attached. In the following sections, the impact limiter is evaluated for the working load limit in the lifting ring and the lifting ring thread engagement. The upper impact limiter design information is:

Secondary Lid Weight WUL= 2541 kg, assume 2700 kg Number of Lifting Rings nr = 3 Dynamic Load Factor DLF = 1.35 2.5.1.3.4.1 Lifting Ring Working Load The lifting rings on the upper impact limiter are used only for lifting when the impact limiter is detached from the cask body; the rings are rendered inoperable by removing the rings from the impact limiter when the cask is assembled. Since the rings are not considered a structural part of the package, they do not need to be designed for the factor of safety against yielding.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Lifting Ring Load WUL x DLF 2700 x 1.35 Pr = = = 1215 kg nr 3 Ring Working Load Limit Pr,max = 3000kg Factor of Safety

, 3000

= = = 2.47 > 1.0 1215 2.5.1.3.4.2 Impact Limiter Thread Engagement The minimum required thread engagement length to prevent impact limiter material failure is determined in accordance with Machinerys Handbook [Ref. 27]. The upper impact limiter is manufactured from ASTM A240 Dual Certified Type 304/304L material. This material is weaker than the M20 lifting ring material (ASTM A-354 Gr. BD), so failure will occur at the root of the upper impact limiter material threads. The minimum required thread engagement length that prevents upper impact limiter material failure is:

Minimum Engagement Length Le = Sbt 2 A b 1

Sct n Ds,min + 0.57735 ( Ds,min E n,max )

2 n Sbt = Bolt External Thread Tensile Strength, MPa Ab = Stress Area of Bolt External Threads, mm² Sct = Cask Internal Thread Tensile Strength, MPa n = Number of threads per millimeter Ds,min = Minimum Major Bolt Diameter, mm En,max = Maximum Pitch Diameter of Internal Thread, mm Solving the equation for Minimum Engagement Length, Le:

Minimum Engagement Length 150,000 x 2 x 0.38 Le =

1 69,000 x x 10.16 x 0.773 x [2 x 10.16 + 0.57735 x (0.773 0.699)]

= 0.73 in = 18.5 mm Where Sbt = 1030 MPa = 150,000 psi Ab = 245.0 mm² = 0.38 in² Robatel Technologies, LLC Page 2-25

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 SLt = 470 MPa = 69,000 psi p = Thread Pitch = 2.5 mm = 0.098 in 1 1 n = = = 10.16 Threads/inch p 0.098 Ds,min = 19.623 mm = 0.773 in En,max = 17.744 mm = 0.699 in The available thread engagement, Lep, is 32 mm. Therefore, the factor of safety is:

32.0

= = 18.5 = 1.73 > 1.0 Therefore, the upper impact limiter lifting ring configuration is acceptable for the required loads.

2.5.1.3.5 Lower Impact Limiter Lifting Evaluation The lower impact limiter is lifted using three of the threaded bolt studs that are utilized to attach the lower limiter to the cask body. As such, it cannot be lifted while attached to the cask body. The lower impact limiter is evaluated for the bolt stresses and for minimum thread engagement in the lower impact limiter during lifting activities. The lower impact limiter design information is:

Lower Impact Limiter Weight WLL = 2448 kg, assume 2600 kg Number of Lifting Rings nr = 3 Dynamic Load Factor DLF = 1.35 Gravitational Acceleration g = 9.81 m/s² 2.5.1.3.5.1 Attachment Bolt Stresses The bolts on the lower impact limiter are only used for lifting when the lower impact limiter is detached from the cask body, and are rendered inoperable by securing them to the cask body as part of the assembled cask. The bolts are therefore not considered to be a structural part of the package with respect to lifting and do not need to be designed for the factor of safety against yielding. Since the arrangement of the cables or straps used to lift the lower impact limiter may vary, the total lifting load is conservatively considered simultaneously in the vertical and horizontal directions.

WLL xDLFxg 2600 x1.35x9.81 Bolt Tension T = = = 11477.7 N nb 3 WLL xDLFxg 2600 x1.35x9.81 Bolt Shear V = = = 11477.7 N nb 3 Bolt Stress Area Ab = 0.000817 m² T 11477.7 kN Bolt Tensile Stress 1 = = = 14048.6 m2 = 14.0 MPa Ab 0.000817 x1000 V 11477.7 kN Bolt Shear Stress = = = 14048.6 m2 = 14.0 MPa Ab 0.000817 x1000 Robatel Technologies, LLC Page 2-26

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1

Maximum Principal Stress p1 = x [1 + 21 + 4 x 2 ]

2 1

= x [14.0 + 14.02 + 4 x 14.02 ] = 22.7 MPa 2

1 Minimum Principal Stress p2 = x [1 21 + 4 x 2 ]

2 1

= x [14.0 14.02 + 4 x 14.02 ] = 8.7 MPa 2

p1 p2 22.7(8.7)

Maximum Shear Stress max = = = 15.7 MPa 2 2 Bolt Yield Stress Sy = 896.3 MPa Allowable Shear Stress Sa = 0.6 x Sy = 537.6 MPa Sa 537.6 Factor of Safety FS = = = 34.2 > 3.0 max 15.7 2.5.1.3.5.2 Lower Impact Limiter Thread Engagement The minimum required thread engagement length to prevent impact limiter material failure is determined in accordance with Machinerys Handbook, 26th Edition [Ref. 27]. Since the constants in the equation assume U.S. customary units, the metric units used in this calculation are converted for determination of the required engagement length. The minimum required thread engagement length that prevents upper impact limiter material failure is:

Minimum Engagement Length Le = Sbt 2 A b 1

Sct n Ds,min + 0.57735 ( Ds,min E n,max )

2 n Sbt = Bolt External Thread Tensile Strength, MPa Ab = Stress Area of Bolt External Threads, mm² Sct = Cask Internal Thread Tensile Strength, MPa n = Number of threads per millimeter Ds,min = Minimum Major Bolt Diameter, mm En,max = Maximum Pitch Diameter of Internal Thread, mm Solving the equation for Minimum Engagement Length, Le:

Minimum Engagement Length 150,000 x 2 x 1.27 Le =

1 69,000 x x 6.35 x 1.396 [ + 0.57735 x (1.396 1.313)]

2 x 6.35

= 1.56 in = 39.5 mm Where Sbt = 1030 MPa = 150,000 psi Robatel Technologies, LLC Page 2-27

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Ab = 817.0 mm² = 1.27 in² SLt = 470 MPa = 69,000 psi p = Thread Pitch = 4.0 mm = 0.157 in 1 1 n = = = 6.35 Threads/inch p 0.157 Ds,min = 35.465 mm = 1.396 in En,max = 33.342 mm = 1.313 in The available thread engagement, Lep, is 75 mm. Therefore, the factor of safety is 75.0

= = 39.5 = 1.90 > 1.0 Therefore, the lower impact limiter lifting ring configuration is acceptable for the required loads.

2.5.2 Tie-down Devices The RT-100 cask utilizes two sets of tie down arms, as shown in Chapter 7, Figure 7.4.4-1. These tie-down arms are welded to two different tie-down plates that in turn are welded to the outer shell of the cask body. Each set of arms on opposite sides of the cask are designed to cross over and securely position the cask, and to absorb the latitudinal, longitudinal and vertical forces required by 10 CFR 71.45 [Ref. 2]. The tie-down arms and plates are a structural part of the package, and must withstand the following loads without impairing the safety of the cask:

o Two (2) times the loaded weight of the cask in the vertical direction o Ten (10) times the loaded weight of the cask in the direction of travel o Five (5) times the loaded weight of the cask transverse to the direction of travel These loads are considered to act simultaneously on the cask and the tie-down arms.

The lifting pockets on the cask body are the only other parts of the cask that could possibly be used to tie down the cask. As such, these pockets are rendered inoperable for tie-down during transport by ensuring that the lift yoke retaining pins are installed in place prior to transport.

2.5.2.1 Tie-down Load Calculation The maximum forces applicable in each of the three loading directions are calculated in this section.

This calculation is accomplished by using the mass of the fully loaded cask along with the gravitational acceleration and the vertical, longitudinal and lateral factors specified in 10 CFR 71.45

[Ref. 2]. The loaded weight of the cask is specified in Chapter 1, Section 1.2.1.2.

Gravitational Acceleration: g = 9.81 m/s2 Cask Mass: Mc = 34696 kg Payload Mass: Mp = 7060 kg Total Mass: M = Mc + Mp = 34696 kg, assume 42000 kg Total Weight: W = Mg = 412.02 kN Vertical Acceleration dv = 2 Robatel Technologies, LLC Page 2-28

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Axial Acceleration da = 10 Transverse Acceleration dL = 5 Vertical Load Py = M x g x dy = 824 kN Axial Load Pa = M x g x da = 4120.2 kN Transverse Load PL = M x g x dL = 2060.1 kN 2.5.2.2 Tie-down Force Calculation The geometric configuration of the tie-down system is designed so that the resultant tie-down arm tensile loads are tangent to the cask surface in order to minimize the effects of out-of-plane stresses in the cask shell. Figure 2.5.2-1 and Figure 2.5.2-2 illustrate the details of the tie-down system geometry. Shear stops are utilized to convert some of the cask loads into turning moments that are restricted by the tie-down arms. As shown on drawing RT PE 1001-1 Rev. F - Robatel Transport Package RT-100 General Assembly Sheet 1/2 (Chapter 1, Appendix 1.4, Attachment 1.4-2), the tie-down arms have slightly different angles in the front and rear of the casks. These differences are summarized in Table 2.5.2-1. The horizontal angles from the cask body to each arm varies from 40o and 44o on one end of the cask and 37o and 41o on the other.

Table 2.5.2-1 Tie-down Arms Horizontal Angles Load Arms in Tension Angles Average Angle L & M (Rear) 44 and 40 42 Longitudinal Q & R (Front) 37 and 41 39 M&R 40 and 41 40.5 Lateral L&Q 44 and 37 40.5 Vertical L, M, Q, R 44, 40, 37, 41 40.5 The analytical model for determining the reaction loads required to prevent rotation and translation of the package due to the 10 CFR 71.45 [Ref. 2] applied loads is shown in Figure 2.5.2-1 and Figure 2.5.2-2. The evaluation is bounded by analyzing the high average angle (42o) caused by longitudinal forces on the tie-down arms on the rear of the cask, and the low average angle (32o) caused by longitudinal forces on the tie-down arms on the front of the cask. The shear stop forces at the bottom of the package are represented by the orthogonal components of a single force vector, S, making an angle of with the global yaxis. The stresses in the members are determined by considering the component loads (10W, 5W, and 2W) individually and superimposing the results. The geometry of the arms has a slight asymmetry so that the tie downs can cross one another; this slight asymmetry is ignored and average dimensions are used for calculation purposes. A detailed force analysis is conducted using the dimensions and notations shown in the figures; other terms are defined below:

W: weight of cask, kN Tx: tensile force in member 2 and 3 resulting from 5W load, kN Ty: tensile force in member 1 and 2 resulting from 10W load, kN Tz: tensile force in each member resulting from 2W load, kN T1,2,3,4: total tensile force in subscripted member, kN Fx: total force in the x direction resulting from 5W load, kN Robatel Technologies, LLC Page 2-29

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Fy: total force in the y direction resulting from 10W load, kN L: Effective length of tie-down arm, i.e. distance between tie-down tangent point and center of tie-down attachment eye, mm The forces are derived in detail in Calculation Package RTL-001-CALC-ST-0202, Rev. 4 [Ref. 34]

and are developed via summing the moments about the center of gravity. A summary of the values calculated using Figure 2.5.2-1 and Figure 2.5.2-2 are provided in Table 2.5.2-2. The maximum calculated forces using these values is provided in Table 2.5.2-3. The results show that the front arms with the lower horizontal angle are subjected to the greater forces. The evaluation of the longitudinal loads on the two front tie-down arms bounds the evaluation of all other load conditions on the cask. The tension calculations and safety margin evaluations contained in the following sections focuses on the front tie-down arms.

weight 412.02 KN R 1293.5 mm r 1050 mm d 1648 mm t 1429 mm L 605 mm 0.514872 rad 0.733038 rad a 352.3409 mm b 391.3142 mm c 297.9163 mm x' 427.9612 mm y' 1093.901 mm z' 1726.916 mm Figure 2.5.2-1 RT-100 Tie-Down Arm Geometry Robatel Technologies, LLC Page 2-30

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.5.2-2 RT-100 Tie-Down Free Body Diagrams Robatel Technologies, LLC Page 2-31

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.5.2-2 Calculated Values for Tie-Down Arms Rear Arms Front Arms o o (44 & 40 )= > 0.733038 rad o o (41 & 37 )= > 0.680678 rad a 351.47 mm 365.61 mm b 390.34 mm 451.49 mm c 297.18 mm 328.69 mm L (616 + 591)/2 = 603.5 mm (682 + 653)/2 = 667.5 mm x 451.13 mm 473.71 mm y 1113.01 mm 1131.16 mm z 1726.18 mm 1757.59 mm (Note: these values calculated using parameters as defined in Figure 2.5.2-1 and Figure 2.5.2-2)

Table 2.5.2-3 Calculated Forces for Tie-Down Arms Rear Arms Front Arms Tx 1361.26 kN 1430.82 kN Ty 1609.56 kN 1571.40 kN Tz 418.36 kN 418.36 kN Tmax 3389.18 kN 3420.58 kN Fxx 474.56 kN 492.68 kN Fyy 2038.07 kN 1994.43 kN Fn 2925.80 kN 2956.73 kN Ff 146.29 kN 147.84 kN Sx 204.57 kN 213.28 kN Sy 953.61 kN 931.10 kN 2.5.2.3 Tie-Down Arm Evaluation The maximum tie-down arm load of 3420.58 kN is determined as described in Section 2.5.2.2 above. This load is applied to the tie-down arm design to ensure that stresses are within allowable limits. As show in the drawings presented in (Chapter 1, Appendix 1.4, Attachments 1.4-2 through 1.4-8) the tie-down arm is reinforced in the portion containing the attachment hole. This reinforcement ensures that the loads in this area of reduced cross-section can be transmitted safely into the rest of the tie-down arm. Stresses for the tie-down arm and its connection to the exterior cask shell are calculated as follows:

Arm Tension Stress at Hole Arm Cross-Sectional Area at Hole, Anet = 11,450 mm2 Arm Tension Stress, net = Tmax / Anet = 298.74 MPa Stress Allowable, allow = 437.2 MPa (@50°C per Table 2.2.1-1)

Factor of Safety, FS = allow / net = 437.2 / 298.74 = 1.46 > 1.0 Arm Bearing Stress at Hole Arm Bearing Area at Hole, Abear= 7,650 mm2 Arm Tension Stress, net = Tmax / Abear = 447.13 MPa Stress Allowable, allow = 1.35 x 437.2 MPa = 590.2 MPa (@50°C per Table 2.2.1-1)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Factor of Safety, FS = allow / net = 590.2 / 447.13 = 1.32 > 1.0 Arm Tear-Out Stress at Hole Arm Tear-out Area, Atear = 18,700 mm2 Arm Tear-out Stress, tear = Tmax / Atear = 182.92 MPa Tear-out Stress Allowable, allow = 0.6 x 437.2 = 262.3 MPa Factor of Safety, FS = allow / tear = 262.3 / 182.92 = 1.43 > 1.0 Arm Tension Stress at Main Cross Section Arm Area, Aarm = 9,100 mm2 Arm Tear-out Stress, arm = Tmax / Aarm = 375.89 MPa Tear-out Stress Allowable, allow = 437.2 MPa Factor of Safety, FS = allow / arm = 437.2 / 375.89 = 1.16 > 1.0 As shown in the summary above, the stresses in the limiting tie-down arm are below the yield stress allowables.

2.5.2.4 Tie-down Arm & Plate Weld Evaluation The stresses in the welds attaching the tie-down arms to the tie-down plates and the plates to the cask body are found by applying the loads from the attachment arms to the weld around the perimeter of the plates. The maximum load on the tie-down arm welds are the sum of the loads in two connecting arms. Thus, from inspection of Figure 2.5.2-2, the maximum tie-down arm load is calculated as follows:

Tie-down Arm Weld Force, Ftotal = 2Tx + Ty +2Tz = 5269.76 kN Weld axial load Fx = Ftotal x (b / L) = 3564.43 kN Weld vertical load Fy = Ftotal x (c / L) = 2594.96 kN Weld transverse load Fz = Ftotal x (a / L) = 2886.42 kN Arm tensile strength: 437.2 MPa Cask tensile strength: 199.3 MPa Weld tensile strength: 450 MPa, weld between tie-down arm and plate [Ref. 34]

420 MPa, weld between tie-down plate and cask [Ref. 34]

The weld length, b, is 1583.36 mm, the weld height d for the tie- down arm plate is the 260 mm height of the arm, and weld height d for the weld between tie- down plate and cask body is 388.03 mm (Calculation Package RTL-001-CALC-ST-0202 Rev. 4 [Ref. 34]). These dimensions and loads are used in the following weld stress calculations.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.5.2.4.1 Tie Down Arm-to-Plate Weld Stress The stresses in the welds attaching the tie-down arm to the tie-down plate are found by applying the weld loads as specified in Section 2.5.2.4. The stresses and allowables are determined as described in Design of Welded Structures [Ref. 25] and Calculation Package RTL-001-CALC-ST-0202, Rev. 4 [Ref. 34].

Y Weld properties are as follows:

Cx b = 1.583 m X X d = 0.260 m d Cy = b/2 = 0.79 m Cx = d/2 = 0.13 m Aw = 2 x b = 3.172 m3/m Y Y Sx = b x d = 0.41 m3/m Cy Sy =b2/3 = 0.84 m3/m X Local Jw = b (3d2 + b2) / 6 = 0.71 m4/m b Coordinates Weld Throat Size = 0.022 m Weld stress is calculated as follows:

ft = (Fz / Aw) + (Mx / Sx) + (My / Sy) = 911.69 kN/m fvy = (Fy / Aw) + ((Mz x Cy) / Jw) = 819.63 kN/m fvx = (Fx / Aw) + ((Mz x Cx) / Jw) = 1125.85 kN/m fw = (ft² + fvy² + f vx² ) 1/2 = 1664.49 kN/m Weld Allowable Stress = 0.6 x Fw x Weld Size x 1000 = 5940 kN/m Weld Metal Factor of Safety, FS = 5940 / 1664.49 = 3.56 > 1.0 Tie-Down Arm Shear Allowable = 0.6 x Fw x Weld Size / 0.7071 x 1000 = 8158 kN/m Tie-Down Arm Factor of Safety, FS = 8158 / 1664.49 = 4.90 > 1.0 2.5.2.4.2 Tie Down Plate-to-Outer Shell Weld Stress The stresses in the welds attaching the tie-down plate to the cask outer shell are found by applying the weld loads as specified in Section 2.5.2.4. The stresses and allowables are determined as described in Design of Welded Structures [Ref. 25] and Calculation Package RTL-001-CALC-ST-0202, Rev. 4 [Ref. 34].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Weld properties are as follows:

Y b = 1.583 m d = 0.388 m Cx Cy = b/2 = 0.79 m Cx = d/2 = 0.19 m X X d

Aw = 2 x b = 3.172 m3/m Sx = b x d = 0.615 m3/m Sy =b2/3 = 0.84 m3/m Y Y

Jw = b (3d2 + b2) / 6 = 0.78 m4/m Cy X

Weld Throat Size = 0.017 m b Local Coordinates Weld stress is calculated as follows:

ft = (Fz / Aw) + (Mx / Sx) + (My / Sy) = 911.69 kN/m fvy = (Fy / Aw) + ((Mz x Cy) / Jw) = 819.63 kN/m fvx = (Fx / Aw) + ((Mz x Cx) / Jw) = 1125.85 kN/m fw = (ft2 + fvy2 + fvx2)1/2 = 1664.49 kN/m Weld Allowable Stress = 0.6 x Fw x Weld Size x 1000 = 4284 kN/m Weld Metal Factor of Safety, FS = 4284 / 1664.49 = 2.57 > 1.0 Outer Shell Shear Allowable = 0.6 x Fw x Weld Size / 0.7071 x 1000 = 2.875 kN/m Outer Shell Factor of Safety, FS = 2875 / 1664.49 = 1.73 > 1.0 2.5.2.5 Tie-Down Evaluation Summary As shown in the previous sections, all components of the tie-down components that are a structural part of the cask maintain positive safety margins when subjected to the simultaneous loadings specified in 10 CFR 71.45 [Ref. 2]. The smallest factor of safety is 1.16 against tie-down arm tension. Under excessive loading, the failure of the tie-down system occurs by yielding in the tie-down arm. This failure does not impair the packages ability to meet other regulatory requirements since the tie-down arms are welded to a plate that is in-turn welded to the cask body. Damage to the tie-down arm does not damage any component integral to the cask body and therefore, does not compromise the cask body shell.

2.6 Normal Conditions of Transport This Section describes the RT-100 evaluation for the normal conditions of transport specified in 10 CFR 71.71[Ref. 2]. The requirements of 10 CFR 71.71 state that the RT-100 shall be structurally adequate for the following normal conditions of transport:

o Heat o Cold o Reduced external pressure o Increased external pressure Robatel Technologies, LLC Page 2-35

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Vibration o Water spray, free drop o Corner drop o Compression, and o Penetration.

During the free drop analyses, the cask impact orientation evaluated is the orientation that inflicts the maximum damage to the cask. Also, the requirements of 10 CFR 71.71 [Ref. 2] specify that the evaluation of the RT-100 for the normal conditions of transport be evaluated at the most unfavorable ambient temperature in the range from -29C to +38C. The normal conditions of transport evaluations presented in this section show that the package satisfies the applicable performance requirements specified in the 10 CFR 71.71 [Ref. 2]. The scale drop testing and analytical analyses demonstrate that there is no decrease in the RT-100 Cask Package effectiveness as follows:

o No loss or dispersal of contents o No structural changes reducing the effectiveness of components required for shielding, for heat transfer, or for maintaining subcriticality or containment o No changes to the package affecting its ability to withstand HAC.

The normal conditions evaluations described in the following sections are performed in accordance with the design criteria and load combinations as identified in Section 2.1.2. Each of the following subsections addresses each normal conditions requirement.

2.6.1 Heat The RT-100 cask body and closure lids are analyzed for structural adequacy in accordance with the thermal evaluation of the RT-100 for the temperatures specified in 10 CFR 71.71(c)(1) [Ref. 2] is presented in Chapter 3. The thermal evaluation demonstrates that the cask component temperatures are maintained within their safe operating ranges for all normal conditions of transport. The following subsections discuss the structural evaluation of the RT-100 using the appropriate component temperatures as determined in Chapter 3.

2.6.1.1 Summary of Pressures and Temperatures The pressures and temperatures occurring in the RT-100 as a result of the 10 CFR 71 [Ref. 2]

normal conditions of transport thermal conditions are an important consideration for the structural evaluations presented in this chapter. The internal pressure induces stresses on the containment system; the temperatures affect the selection of temperature-dependent material properties as well as, the internal pressures that occur as a result of the ambient temperatures and solar insolation specified in 10 CFR 71.71 [Ref.2]. The material properties utilized are based on the maximum calculate temperatures of each component or higher temperatures which are conservative. For all NCT structural analyses and finite element analyses, a minimum normal operating pressure of 35 psig is uniformly applied to the interior of the inner shell [Ref. 35, Section A.4.]. Note that a reference to the assumed minimum normal operating pressure may appear in psia unit in various sections of the SAR.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The maximum normal operating pressure evaluation for the RT-100 is presented in Chapter, 3 Section 3.3.2. As described in this section, the calculated maximum pressure for normal conditions is 182.71 kPa (26.5 psia). For conservatism, the structural evaluations involving internal pressure use a maximum normal operating condition pressure of 342.7 kPa (49.7 psia).

The maximum component temperatures in the RT-100 for normal conditions are presented in Chapter 3, Table 3.1.3-1 RT-100 Maximum Normal Condition Temperature Summary (Found in Chapter 3). The temperatures are utilized to determine the stress allowables used in the structural evaluation for the normal conditions of transport.

2.6.1.2 Differential Thermal Expansion As shown in Chapter 3, Table 3.1.3-1, the temperatures of the components of the cask differ by only a few degrees under normal conditions of transport thermal ambient conditions. This difference is due in part to the relatively low decay heat of the contents. The RT-100 is evaluated for differential thermal expansion as described in Section 2.6.7 in combination with normal pressure and inertial loads under the following conditions:

o Ambient temperature, 38°C o Initial temperature, 38°C o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o Steady-state solar insolation o Internal heat load as a uniform heat flux, 13.04 W/m2 2.6.1.3 Stress Calculations Regulatory Guide 7.6 [Ref. 4] requires that the range of primary plus secondary stress intensities during normal conditions of transport be less than 3.0 Sm. To evaluate this condition, the range of primary plus secondary stresses for the combined normal events (including heat, cold, normal operating pressure, 0.3-m end drop, and 0.3-m side drop conditions) are analyzed using the finite element model presented in 2.6.7.2.

2.6.1.4 Comparison with Allowable Stresses The combined stress results are presented in Tables 2.6.7-1 and 2.6.7-2. Since the margins of safety are all positive, the RT-100, therefore, satisfies the requirements of 10 CFR 71.71(c)(1) [Ref. 2] for the heat (normal transport) condition.

2.6.2 Cold The RT-100 cask body and closure lids are analyzed for structural adequacy in accordance with the thermal evaluation of the RT-100 for the temperatures specified in 10 CFR 71.71(c)(2) [Ref. 2] is presented in Chapter 3. The thermal evaluation demonstrates that the RT-100 component temperatures are maintained within their safe operating ranges for all normal conditions of transport. Using the same methodology presented in Section 2.6.1, the RT-100 is evaluated for cold conditions. The following thermal case is used to calculate the thermal stress under cold conditions:

o Ambient temperature, -40°C Robatel Technologies, LLC Page 2-37

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Initial temperature, -40°C o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o No solar insolation, in shade o Internal heat load as a uniform heat flux, 13.04 W/m2 The combined stress results are presented in Tables 2.6.7-1 and 2.6.7-2. Since the margins of safety are all positive, the RT-100, therefore, satisfies the requirements of 10 CFR 71.71(c)(2) [Ref. 2] for the cold (normal transport) condition.

2.6.3 Reduced External Pressure The drop in atmospheric pressure to 24 kPa (3.5 psia), as specified in 10 CFR 71.71(c)(3) [Ref. 2],

effectively results in an additional internal pressure in the cask of 77 kPa (11.2 psig). This additional pressure has a negligible effect on the RT-100 because, in Section 2.6.1.1, the cask is analyzed for a normal transport conditions internal pressure of 241 kPa (35 psig). Maximum internal pressure is included in combination with internal loads (see Tables 2.6.7-1 and 2.6.7-2).

Since the margins of safety are all positive, the RT-100 satisfies the requirements of 10 CFR 71.71(c)(3) for reduced external pressure.

2.6.4 Increased External Pressure An increased external pressure of 20 psia (5.3 psig external pressure), as specified in 10 CFR 71.71(c)(4) [Ref. 2], has a negligible effect on the RT-100 because of the thick outer shell and end closures of the cask. Section 2.6.7 addresses many different loading cases which exceed these prescribed external pressure requirements. Therefore, the requirements of 10 CFR 71.71(c)(4)

[Ref. 4] are satisfied.

2.6.5 Vibration 10 CFR 71.71 (c)(5) [Ref.4] requires that vibration normally incident to transport be evaluated.

The RT-100 package consists of think section materials that are unaffected by vibration normally incident to transport, such as over the road vibrations.

2.6.5.1 Vibration Evaluation of the RT-100 Cask Primary Lid Bolts The RT-100 may be subjected to a cycle range typically associated with high-cycle fatigue (> 108 cycles). Therefore, the endurance limit of the material for the high cycle fatigue can be approximated by using a 60% reduction, rh, of the ultimate tensile strength (AISC [Ref. 26]) with an additional 10% reduction rg, for the connection surface (Machinerys Handbook [Ref. 27]).

Thus the endurance limit for the material is:

Sa = (1 rh ) (1 rg ) Sub where:

Sub = Bolt Ultimate Stress

= 1030 MPa (ASTM A354 Grade B, Table 2.2.1-3)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Sa = (1 0.60) (1 0.10) 1030

= 370.8 MPa NUREG-0128 [Ref. 30] gives the following RMS vibration load factors for the road travel:

fv = Vertical Vibration Load Factor

= 0.52 fL = Longitudinal Vibration Load Factor

= 0.27 ft = Transverse Vibration Load Factor

= 0.19 The RT-100 is transported in the vertical orientation. The cask lid is subjected to vibration in the vertical direction. A notch factor, fN, of 3.0 is used and is conservative (AISC [Ref. 26]). The vibration stress in the bolts is:

Fb f N sy =

Ab where:

Fb = Bolt Force due to Vibration f v WLp g

=

Nb Ab = Bolt Stress Area

= 1470 mm2 [Ref. 27]

WLp = Cask Lid Weight

= 3648 kg, use 3650 kg Nb = Number of Bolts

32 0.52 3650 9.81 1 kN Fb

32 1000 N

0.58 kN 0.58 3.0 1 MPa sv

0.001470 1000 kN 2 m

= 1.19 MPa << Sa = 370.8 MPa Robatel Technologies, LLC Page 2-39

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Since the stress in the bolts is well below the endurance limit of the material, the primary lid bolts are not subjected to transportation-related fatigue damage during their service life.

The maximum shock loading coefficient for the three orthogonal directions is specified as 2.9 (NUREG-0128 [Ref. 30]). The RT-100 primary lid is subjected to shock loading during transport.

The primary lid closure bolts are shown to withstand a 125g impact load (Section 2.13.3.3), which is much larger than the 2.9W shock loading during transport. Therefore, the primary lid closure bolts are acceptable for shock loading by comparison.

2.6.5.2 Vibration Evaluation of the RT-100 Cask Secondary Lid Bolts Per Section 2.6.5.1, the components of the package are in the high-cycle fatigue range (> 108 cycles).

The endurance limit of the material for the high cycle fatigue for the secondary lid bolts is the same as for the primary lid bolts. The RT-100 lid is subjected to vibration in the vertical direction. A notch factor, fN, of 3.0 is used and is conservative (AISC [Ref. 26]). The vibration stress in the bolts is:

Fb f N sv =

Ab where:

Fb = Bolt Force due to Vibration f v WLp g

=

Nb Ab = Bolt Stress Area

= 817 mm2 [Ref. 27]

W Ls = Cask Lid Weight

= 857 kg Nb = Number of Bolts

18 All other quantities are defined in Section 2.6.5.1 0.52 857 9.81 1 kN Fb

18 1000 N

0.24 kN 0.24 3.0 1 MPa sv

0.000817 1000 kN 2 m

= 0.89 MPa << Sa = 370.8 MPa Since the stress in the bolts is well below the endurance limit of the material, the secondary lid bolts Robatel Technologies, LLC Page 2-40

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 are not subjected to transportation-related fatigue damage during their service life.

The maximum shock loading coefficient for the three orthogonal directions is specified as 2.9 (NUREG-0128 [Ref. 30]). The cask primary lid is subjected to shock loading during transport.

The secondary lid closure bolts have been show to withstand a 125g impact load (Section 2.12.4.1),

which is much larger than the 2.9W shock loading during transport. Therefore, the secondary lid closure bolts are acceptable for shock loading by comparison.

The RT-100 satisfies the requirements for normal vibration incident to transport as required by 10 CFR 71.71(c)(5) [Ref. 2].

2.6.6 Water Spray Water causes negligible corrosion of the stainless shell of the RT-100. The cask contents are protected in the sealed cavity. A water spray as specified in 10 CFR 71.71(c)(6) [Ref. 2] has no adverse impact on the package. The cask surface temperature specified during the water spray is between 38C and -29C. Consequently, the induced thermal stress in the cask components is less than the thermal stresses that occur during the extreme temperature conditions for normal transport.

Therefore, the requirements of 10 CFR 71.71(c)(6) [Ref. 2] are satisfied.

2.6.7 Free Drop The RT-100 is shown to meet the free drop requirements of 10 CFR 71.71 [Ref. 2] through a combination of classic calculations, finite element analyses and scale model drop testing (RTL-001-CALC-ST-0402, Rev. 4 [Ref. 35]). The evaluations include the qualification of the RT-100 cover bolt design for the combined effects of free drop impact force, internal pressures, thermal stress, O-ring compression force, and bolt preload following the methodology of NUREG/CR-6007

[Ref. 10] (Appendix 2.13). The combined effects of inertial loads, internal pressures, and thermal stress are considered for packaging components.

2.6.7.1 Methodology The RT-100 is designed in accordance with Regulatory Guide 7.6 [Ref. 4]. The design criteria for NCT and HAC are presented in Table 2.1.2-2. Load combinations for the structural analysis of shipping casks for radioactive materials are defined by Regulatory Guide 7.8 [Ref. 3]. The load combinations for all normal and accident conditions and corresponding ASME service levels are shown in Table 2.1.2-1. Material properties used in this evaluation are presented in Section 2.2.1.

Stress intensities caused by pressure, thermal expansion, and mechanical loads are combined before comparing to ASME,Section III, Subsection ND [Ref.7] stress allowables, which are listed in Table 2.2.1-3.

2.6.7.2 Finite Element Analysis The finite element code ANSYS [Ref. 28] is used to generate a three-dimensional model of the RT-100 and to determine its response to normal conditions of transport (NCT) and hypothetical accident conditions (HAC) (Section 2.7.1). Specifically, a one-half (180°) 3D model of the RT-100 inner and outer shells, outer and inner lids, bottom plate and lead shields is constructed using ANSYS [Ref. 28] solid elements. The interaction between components is modeled using gap elements. Stability of the model is assured by using weak springs. Boundary conditions are applied Robatel Technologies, LLC Page 2-41

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 to the model simulating the loading conditions the cask will experience during normal and accident transport conditions. Pressure loads are applied to the cask inner shell to simulate bounding contents loads and internal pressurization. Thermal stresses are calculated using input temperatures from the NCT thermal analyses. Bolt preloads are applied to represent the bolt torque at the time the cask is readied for shipment. Post-processing is accomplished by linearizing the stress across locations where maximum stresses are calculated. The analyses assume linear elastic behavior of the cask. Therefore, calculated stress intensities are compared to appropriate allowables (Table 2.2.1-1) and the margin of safety is calculated.

2.6.7.2.1 Model Description Finite element analysis methods are used to perform the stress evaluation of the RT-100 for normal and accident free drop conditions. Each drop condition is analyzed using a three- dimensional finite element model using the computational modeling software ANSYS [Ref. 28]. Figure 2.6.7-1 shows the major components of the RT-100 represented in the model including the inner and outer shells, flange, bottom plate, primary and secondary lids, and closure bolts.

As shown in Figure 2.6.7-1, the model (which corresponds to half (180°) of the cask body) is generated by de-featuring the SolidWorks solid model used to develop the manufacturing drawings and exporting the model to a .STEP file format. The .STEP file is imported directly into ANSYS [Ref.28] where the finite element model is developed following the guidance presented in ISG-21 [Ref. 53]. The resulting finite element model of the cask body is represented using solid elements, contact elements, mass elements and spring/damper elements (Figure 2.6.7-2).

The solid portion of the model is constructed using ANSYS solid (SOLID185) elements. Surface-to-surface contact elements are used to simulate the interaction between adjacent components.

Specifically, contact between the cask shells and lead shielding are modeled using CONTAC174/TARGE170 surface-to-surface contact elements with zero friction, which allows the lead to float between the inner and outer shells. Contact elements are also used to bond dissimilarly meshed components. To simulate the impact limiters, the interaction between the cask body and impact limiters is modeled using CONTAC52 gap elements (Figure 2.6.7-3), which acts as a compression only element. The size of the CONTAC52 gaps is determined from nominal dimensions between the impact limiter and cask body. Spring elements (COMBIN14) are inserted automatically during the solution to help stabilize the model. ANSYS [Ref. 28] assigns low spring stiffness so their presence does not adversely affect the accuracy of the solution.

Finite element model verification and mesh density study are presented in Appendix A.4 of Calculation Package RTL-001-CALC-ST-0402, Rev. 4 [Ref. 35]. During the development of the finite element model each part and interface was considered on an individual basis. The RT-100 outer shell was meshed using the sweep method and the element size was varied until there was a sufficient number of elements across the shell thickness. The element ratio was reviewed to ensure adequate results. To test a component, in this case the outer shell, the ends were fixed and a pressure load was applied to the inner surface and a solution was obtained. If a singularity or discontinuity was noted, the mesh was refined until uniform results were obtained. As a second check, a hand calculation was performed on to ensure that the stress calculated by ANSYS [Ref. 28] is giving expected results. Hoop stresses were also calculated and compared to the results. As the model was developed the same philosophy was applied to the intersection of the shell and bottom plate. Using Robatel Technologies, LLC Page 2-42

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Roark's equations (Roarks Formulas for Stress and Strain [Ref. 29]), the interface stress was checked to ensure the bending stress was in the expected range.

The choice of element type was evaluated by running a series of sensitivity studies. For this case, a high order 8-node brick element was chosen over brick element with mid-side nodes. This choice was made because of the relatively thin section of the RT-100 shell versus the length, which made it possible to increase the total number of elements without compromising the run time performance. Several cases were run to vary the total mesh density to see how the stress results varied versus performance of the model. In the extreme case, an overly dense mesh produced excessively long run times and un-converged solutions. Models with low mesh densities that were too low resulted in unrealistic stress results. After numerous runs a balance was found between consistent results and model performance with variations in stress results of less than 1% when comparing high mesh densities to adequate mesh densities. Therefore, it was concluded that the cask model was a quality model and met the intent of ISG-21 [Ref. 53].

At the time the analyses were performed, analyses were generally compared to models previously generated for other 10 CFR 71 [Ref. 2] cask designs. The results of the RT-100 cask analysis are consistent with these previous designs and where peak stress are expected. Additionally, confirmatory scale model testing of the RT-100 demonstrated that the methods used to calculate the cask accelerations and impact limiter deformation are consistent with the drop test results.

Therefore, the inertial loads applied to the cask body are conservative.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-1 RT-100 Solid Model Robatel Technologies, LLC Page 2-44

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-2 RT-100 Finite Element Model Robatel Technologies, LLC Page 2-45

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-3 Gap Elements Used to Represent the Impact Limiters for Side and End Drop Configurations Robatel Technologies, LLC Page 2-46

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.6.7.2.2 Boundary Conditions Boundary conditions are applied to the model to simulate the loading conditions the RT-100 experiences during NCT and HAC. The five categories of cask loading considered in the free drop event are closure lid bolt preload, internal pressure load, thermal load, inertial body load and displacement.

o Closure Lid Bolt Preload: The required total bolt preloads on the cask outer and inner lid bolts are 130.6 kN and 72.2 kN, respectively (10). To apply the bolt preload ANSYS

[Ref. 28] pre-tension elements (PRETS179) are used to define the 3-D pre-tension section within the meshed bolt. The PRETS179 element uses a single translation degree of freedom to define pretension direction (Figure 2.6.7-4). The pretension Section is modeled by a set of pretension elements defined by the bolt shaft.

o Pressure Loading: A pressure of 241 kPa (35 psig) is used to envelope the maximum normal operating pressure for all impact loadings considered (Calculation Package RTL-001-CALC-TH-0102, Rev. 6 [Ref, 42]). For accident conditions, a pressure value of 588 kPa (85.3 psig) is used to represent the pressure experienced during fire conditions (Calculation Package RTL-001-CALC-TH-0202, Rev. 6 [Ref. 43]). The internal pressure load is applied as an equivalent static pressure load uniformly applied on the interior surface of the cask.

o Pressure loading contentscask end drop: For the end drop analyses, the content weight is assumed to be uniformly distributed on the cask end and over an area determined by the inside diameter of the RT-100. Therefore, one-half the contents weight of 6,804 kg (15,000 lb) is applied to the cask inner shell bottom plate. The contents pressure load is multiplied by the appropriate g-load to accurately represent the 304.8 mm (1-foot) and 9144 mm (30-foot) end drop. The pressure value is conservatively multiplied by 1.05 to account for the difference between the solid model surface and the tessellated area of the element mesh.

o Pressure loading contentsside drop: For the side drop condition, the contact area between the contents and the cask cavity is approximately 180° (90° on each side of the drop centerline). The inertial load produced by the 6,804 kg (15,000 lb) contents weight is represented as an equivalent static pressure applied on the interior surface of the RT-100. The pressure is uniformly distributed along the cavity length and is varied in the circumferential direction as a cosine distribution. The pressure value is conservatively multiplied by 1.05 to account for the difference between the solid model surface and the tessellated area of the element mesh. The maximum pressure occurs at the impact centerline; the pressure decreases to zero at locations that are 90 either side of the impact centerline, as illustrated in Figure 2.6.7-5. The following formula is used to determine the contents pressures for the side drop analyses, which vary around the circumference.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 This method uses a summation scheme to approximate the integration of the cosine-shaped pressure distribution:

P 18 F total = max A i cos (i ) cos (i )

i =1 F total = 6,804/2 kg Where Pmax = maximum pressure (at impact centerline) i = average angle of subtended arc of ith element measured from centerline at point of impact to obtain vertical component of pressure i = ith circumferential sector i = normalized angle to peak at 0o and to be zero at 90o Ai = ith circumferential area over which the pressure is applied Gap elements are defined at both ends of the cask to simulate the pressure applied by the impact limiters during side drop conditions. This is accomplished by defining the gap stiffness as a cosine function from a maximum value 175 x 106 N/m (1 x 106 1b/in) at the center line to 15.3 x 106 N/m (87,156 1b/in) at 85o from the center line of impact, and a minimal value 175 x 10 3 N/m (100 1b/in) from 90o to 180o. The load distribution that results from the crushing of the impact limiter is shown in Figure 2.6.7-3.

o Thermal: Four bounding load cases are analyzed to evaluate the RT-100 for the range of temperature and solar insolation conditions specified in 10 CFR 71.71 [Ref. 2] for normal conditions that bound the load combinations presented in Regulatory Guide 7.8

[Ref. 3]. To determine the worst-case conditions, an additional study run evaluates extreme cold (-40°C) without internal heat load. The cases that result in the worst-case thermal expansion are as follow:

Condition 1 - Hot Case 1:

a. Ambient temperature, 38oC

b. Initial temperature, 38oC

c. Heat transfer to ambient by natural convection, still air

d. Heat transfer to ambient by radiation

e. Steady-state Solar insolation

f. Internal heat load as a uniform heat flux, 13.04 W/m2 Condition 2 - Hot Case 2:

a. Ambient temperature, 38oC

b. Initial temperature, 38oC

c. Heat transfer to ambient by natural convection, still air

d. Heat transfer to ambient by radiation Robatel Technologies, LLC Page 2-48

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

e. No solar insolation, in shade

f. Internal heat load as uniform heat flux, 13.04 W/m2 Condition 3 - Cold Case 1:

a. Ambient temperature, -40°C

b. Initial temperature, -40°C

c. Heat transfer to ambient by natural convection, still air

d. Heat transfer to ambient by radiation

e. No solar insolation, in shade

f. Internal heat load as a uniform heat flux, 13.04 W/m2 Condition 4 - Cold Case 2:

a. Ambient temperature, -29°C

b. Initial temperature, -29°C

c. Heat transfer to ambient by natural convection, still air

d. Heat transfer to ambient by radiation

e. No solar insolation

f. Internal heat load as a uniform heat flux, 13.04 W/m2 Heat Conditions 1 and 3 bound the differential the worst case thermal expansion between dissimilar materials. Therefore, Heat Conditions 2 and 4 are not considered.

The cask temperature distributions calculated for Conditions 1 and 3 are used as inputs to the ANSYS [Ref. 28] analyses. The ANSYS analyses determine the stresses arising from the thermal expansion of the cask from its initial 21°C condition, including the effects of the differential thermal growth within the components; these effects are a result of the temperature difference across the cask walls. The cask temperature distributions are also used to determine the values of the temperature-dependent material properties.

The temperatures for the structural analysis are obtained from the results file and database file of the thermal analysis by writing the results to an ASCII file using the ANSYS BFINT command.

Nodes for the structural model are transferred to the same coordinate system as used by the thermal run and the thermal results are interpolated for each thermal condition.

o Inertial body load: The inertial effects, which occur during impact, are represented by equivalent static forces, in accordance with the D'Alembert's principle. The inertial body load includes the weight of the empty cask and the weight of the cavity contents.

Accelerations are calculated in Appendix 2.13. An acceleration of 44g and 52g are applied to the model to simulate end drop and side drop conditions, respectively. The inertial load is applied to the cask body using the ANSYS ACEL command equivalent to the normal and accident conditions accelerations corresponding to the 0.3 meter and Robatel Technologies, LLC Page 2-49

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 9 meter drop cases. Since the lead shield is attached to the steel shells with frictionless contact elements, the lead represents the largest physical load applied to the cask structure.

o Displacement boundary conditions: Displacement boundary conditions are applied to enforce symmetry at the cut boundary of the 3D model. All nodes on the symmetry plane are fixed in the UZ direction. The overall model is stabilized by the gap elements (CONTAC52) that represent the impact limiter, which are connected to the cask body with the outer nodes or ground nodes representing the impact limiter fixed.

Figure 2.6.7-4 Bolt Pre-load Using ANSYS Pre-Tension Elements (PRETS179)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-5 Pressure Distribution Used to Simulate the Contents Robatel Technologies, LLC Page 2-51

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.6.7.3 Side Drop In accordance with the requirements of 10 CFR 71.71 [Ref. 2], the RT-100 is structurally evaluated for the normal condition of transport 0.3 meter side-drop. During the 0.3 meter side-drop event, the cask (equipped with an impact limiter over each end) falls a distance of 0.3 meter onto a flat, unyielding, horizontal surface. The cask strikes the surface in a horizontal position, thereby resulting in a side impact of the cask. The types of loading involved in a side-drop event are lid closure bolt preload, internal pressure load, thermal load, and inertial body load.

Stress results for the 0.3 meter side drop combined loading conditions discussed previously are documented in Table 2.6.7-1. The table documents the primary membrane (Pm), primary membrane plus primary bending (Pm+Pb), primary membrane plus primary bending plus secondary peak stress (Pm+Pb+Q) in accordance with the criteria presented in Regulatory Guide 7.6 [Ref. 4].

As shown in Table 2.6.7-1, the margins of safety are positive when compared to the stress intensity for each category. The most critically stressed component in the system is the inner lid. The minimum margin of safety is found to be +0.8 for primary membrane plus bending stress intensity.

The locations of the critical sections correspond to the maximum stress location shown in Figures 2.6.7.3-1 through 2.6.7.3-11. The minimum margin of safety for primary plus secondary stress intensity is +1.5.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.6.7-1 NCT Side Drop Stress Summary RG 7.6 Margin of Component and Stress ANSYS Results (MPa) Allowable Safety (1)

Stress State Location S1 S2 S3 SINT Stress Pm 5.0 -3.8 -31.6 36.6 138 2.8 Inside 5.3 -3.8 -31.4 36.7 207 4.6 Pm + P b Center 5.0 -3.8 -31.6 36.6 207 4.7 INNER SHELL Outside 4.7 -3.8 -31.8 36.5 207 4.7 Inside 5.3 -3.8 -31.4 36.7 414 10.3 Hot Pm + Pb + Q Center 5.0 -3.8 -31.6 36.6 414 10.3 Outside 4.7 -3.8 -31.8 36.5 414 10.3 Inside 5.3 -3.8 -31.4 36.7 414 10.3 Cold Pm + Pb + Q Center 5.0 -3.8 -31.6 36.6 414 10.3 Outside 4.7 -3.8 -31.8 36.5 414 10.3 Pm 4.3 -3.8 -32.3 36.6 138 2.8 Inside 4.4 -3.8 -32.2 36.5 207 4.7 Pm + Pb Center 4.3 -3.8 -32.3 36.6 207 4.7 OUTER SHELL Outside 4.2 -3.9 -32.5 36.7 207 4.6 Inside 4.4 -3.8 -32.2 36.5 414 10.3 Hot Pm + Pb + Q Center 4.3 -3.8 -32.3 36.6 414 10.3 Outside 4.2 -3.9 -32.5 36.7 414 10.3 Inside 4.4 -3.8 -32.2 36.5 414 10.3 Cold Pm + Pb + Q Center 4.3 -3.8 -32.3 36.6 414 10.3 Outside 4.2 -3.9 -32.5 36.7 414 10.3 Pm 4.1 -3.9 -32.9 37.0 138 2.7 Inside 4.1 -3.9 -32.7 36.8 207 4.6 Pm + P b Center 4.1 -3.9 -32.9 37.0 207 4.6 Outside 4.1 -4.0 -33.0 37.1 207 4.6 FLANGE Inside 4.1 -3.9 -32.7 36.8 414 10.2 Hot Pm + Pb + Q Center 4.1 -3.9 -32.9 37.0 414 10.2 Outside 4.1 -4.0 -33.0 37.1 414 10.2 Inside 4.1 -3.9 -32.7 36.8 414 10.2 Cold Pm + Pb + Q Center 4.1 -3.9 -32.9 37.0 414 10.2 Outside 4.1 -4.0 -33.0 37.1 414 10.2 Pm 18.4 -0.3 -18.4 36.8 138 2.7 Inside 51.6 9.5 7.4 44.3 207 3.7 Pm + P b Center 18.4 -0.3 -18.4 36.8 207 4.6 OUTER LID Outside -8.9 -12.7 -47.7 38.8 207 4.3 Inside 62.8 -15.8 -41.9 104.7 414 3.0 Hot Pm + Pb + Q Center 11.4 -12.5 -39.4 50.8 414 7.1 Outside 12.9 -2.4 -41.7 54.5 414 6.6 Inside 116.0 61.8 27.6 88.4 414 3.7 Cold Pm + Pb + Q Center 30.1 5.4 -17.7 47.8 414 7.7 Outside -4.4 -13.7 -55.0 50.7 414 7.2 Pm -1.5 -2.6 -56.9 55.4 138 1.5 Inside -4.2 -19.9 -121.3 117.1 207 0.8 Pm + P b Center -1.5 -2.6 -56.9 55.4 207 2.7 INNER LID Outside 15.9 7.2 0.3 15.7 207 12.2 Inside 2.4 -31.7 -161.7 164.1 414 1.5 Hot Pm + Pb + Q Center 15.2 2.8 -58.4 73.6 414 4.6 Outside 13.5 -5.2 -23.7 37.2 414 10.1 Inside -8.8 -28.7 -148.7 140.0 414 2.0 Cold Pm + Pb + Q Center 4.1 -0.2 -58.8 62.9 414 5.6 Outside 19.5 4.7 -6.9 26.4 414 14.7 Note: (1) The margin of safety is the ratio of Allowable Stress and the Stress Intensity (SINT) minus 1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-6 RT-100 NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-54

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-7 RT-100 Inner Shell NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-55

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-8 RT-100 Outer Shell NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-56

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-9 RT-100 Flange NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-57

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-10 RT-100 Outer Lid NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-58

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-11 RT-100 Inner Lid NCT Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-59

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.6.7.4 End Drop In accordance with the requirements of 10 CFR 71.71 [Ref. 2], the Universal Transport Cask is structurally evaluated for the normal condition of transport 0.3 m end-drop. In this event, the cask (equipped with an impact limiter over each end) falls a distance of 0.3 m onto a flat, unyielding, horizontal surface. The cask strikes the surface in a vertical position; consequently, an end impact on the bottom end or top end of the cask occurs.

As discussed previously, stress results for the 1-ft top and bottom-end drop combined loading conditions are documented in Table 2.6.7-2. The table documents the primary membrane (Pm),

primary membrane plus primary bending (Pm+Pb), primary membrane plus primary bending plus secondary peak stress (Pm+Pb+Q) in accordance with the criteria presented in Regulatory Guide 7.6

[Ref. 4].

As shown in the Table 2.6.7-2, the margins of safety for the primary stress intensity category are positive for all of the 0.3 m top-end drop conditions. The most critically stressed component in the system is the cask flange region due to the bending of the flange due to the inertial load imposed by the cask lids. The minimum margin of safety is found to be +2.4 for primary membrane plus bending stress intensity. The locations of the critical sections correspond to the maximum stress location shown in Figure 2.6.7-12 through Figure 2.6.7-17. The minimum margin of safety for primary plus secondary stress intensity is +0.2.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.6.7-2 NCT End Drop Stress Summary RG 7.6 Component and Stress ANSYS Results (MPa) Margin of Allowable Stress State Location Safety (1)

S1 S2 S3 SINT Stress Pm 2.7 1.2 -7.8 10.5 138 12.1 Inside 2.7 2.0 -12.2 14.9 207 12.9 Pm + Pb Center 2.7 1.2 -7.8 10.5 207 18.7 INNER SHELL Outside 2.9 0.2 -3.6 6.6 207 30.5 Inside 2.7 2.0 -12.2 14.9 414 26.8 Hot Pm + Pb + Q Center 2.7 1.2 -7.8 10.5 414 38.3 Outside 2.9 0.2 -3.6 6.6 414 61.9 Inside 2.7 2.0 -12.2 14.9 414 26.8 Cold Pm + Pb + Q Center 2.7 1.2 -7.8 10.5 414 38.3 Outside 2.9 0.2 -3.6 6.6 414 61.9 Pm 6.5 -0.9 -3.4 9.9 138 12.9 Inside 7.5 1.0 -2.7 10.2 207 19.3 Pm + Pb Center 6.5 -0.9 -3.4 9.9 207 19.9 OUTER SHELL Outside 6.9 0.7 -9.0 15.9 207 12.0 Inside 113.3 39.9 -63.2 176.5 414 1.3 Hot Pm + Pb + Q Center 22.5 -10.9 -16.7 39.2 414 9.5 Outside 25.4 0.5 -33.5 58.9 414 6.0 Inside 10.7 0.5 -4.5 15.3 414 26.1 Cold Pm + Pb + Q Center 18.7 5.7 -4.7 23.5 414 16.6 Outside 10.4 2.4 -9.5 19.9 414 19.8 Pm 5.9 1.5 -12.3 18.1 138 6.6 Inside 0.1 -3.3 -19.5 19.6 207 9.5 Pm + Pb Center 5.9 1.5 -12.3 18.1 207 10.4 Outside 20.1 6.3 -13.6 33.7 207 5.1 FLANGE Inside 48.0 24.1 -219.4 267.4 414 0.5 Hot Pm + Pb + Q Center 12.9 -5.7 -23.8 36.6 414 10.3 Outside 74.0 34.2 -53.9 127.9 414 2.2 Inside 32.8 -42.6 -105.1 137.9 414 2.0 Cold Pm + Pb + Q Center 14.2 2.1 -24.1 38.3 414 9.8 Outside 92.7 71.4 -36.7 129.4 414 2.2 Pm -0.9 -4.0 -14.6 13.7 138 9.1 Inside -7.7 -17.0 -52.6 45.0 207 3.6 Pm + Pb Center -0.9 -4.0 -14.6 13.7 207 14.1 OUTER LID Outside 24.2 9.0 5.1 19.0 207 9.9 Inside 280.5 36.7 -55.4 336.0 414 0.2 Hot Pm + Pb + Q Center 35.3 20.9 -4.7 40.0 414 9.3 Outside 41.6 16.7 -56.7 98.3 414 3.2 Inside -35.0 -71.0 -163.6 128.5 414 2.2 Cold Pm + Pb + Q Center 14.0 4.5 -14.8 28.8 414 13.4 Outside 21.6 -0.3 -22.2 43.8 414 8.4 Robatel Technologies, LLC Page 2-61

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.6.7-2 (Continued)

RG 7.6 Component and Stress ANSYS Results (MPa) Margin of Allowable Stress State Location Safety (1)

S1 S2 S3 SINT Stress Pm 5.7 -2.3 -35.4 41.1 138 2.4 Inside -6.5 -10.3 -67.7 61.3 207 2.4 Pm + Pb Center 5.7 -2.3 -35.4 41.1 207 4.0 Outside 20.8 6.0 -6.5 27.3 207 6.6 Inside -14.6 -27.5 -112.1 97.5 414 3.2 INN Hot Pm + Pb + Q Center 28.9 11.0 -26.3 55.2 414 6.5 Outside 18.9 -8.7 -36.5 55.3 414 6.5 Inside -18.9 -23.7 -93.0 74.1 414 4.6 Cold Pm + Pb + Q Center 9.7 -1.3 -39.2 49.0 414 7.4 Outside 23.4 3.1 -13.5 36.8 414 10.2 Note: The margin of safety is the ratio of the Allowable Stress and the Stress Intensity (SINT) minus 1.

Figure 2.6.7-12 RT-100 NCT Bottom Drop Stress Intensity Results Robatel Technologies, LLC Page 2-62

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-13 RT-100 Inner Shell NCT End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-63

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-14 RT-100 Outer Shell NCT End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-64

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-15 RT-100 Flange NCT End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-65

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-16 RT-100 Outer Lid NCT End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-66

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.6.7-17 RT-100 Inner Lid NCT End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-67

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.6.8 Corner Drop The RT-100 is composed of materials other than fiberboard or wood. Also, the weight of the RT-100 exceeds 100 kg. According to 10 CFR 71.71(c)(8) [Ref. 2], the corner drop test is not applicable to the RT-100.

2.6.9 Compression According to 10 CFR 71.71(c)(9) [Ref. 2], the compression test is not applicable to the RT-100 because the package weight is greater than 5,000 kg.

2.6.10 Penetration According to 10 CFR 71.71(c)(10) [Ref. 2], a penetration test involving a 13-lb (6-kg) penetration cylinder dropped from a height of 1 m is required for evaluation of packages during normal conditions of transport. However, Regulatory Guide 7.8 [Ref. 3] states that the penetration test of 10 CFR 71.71 [Ref. 2] is not considered by the NRC staff to have structural significance for large shipping casks (except for unprotected valves and rupture disks) and is not considered as a general requirement. A penetration test is not performed since the RT-100 has no unprotected valves or rupture disks that could be affected by normal conditions of transport.

2.7 Hypothetical Accident Conditions The RT-100 Cask meets the standards specified in 10 CFR 71.51 [Ref. 2] when subjected to the conditions and tests specified in 10 CFR 71.73 [Ref. 2] for hypothetical accidents. In accordance with 10 CFR 71.73 [Ref. 2], the RT-100 is structurally evaluated for hypothetical accident scenarios of free drop, puncture, fire, crush, and water immersion. In the free-drop and puncture analyses, the cask impact orientation evaluated is the one that inflicts the maximum damage to the cask. The most unfavorable ambient temperature condition during operation in the range from -40°C to 38°C is assumed. The following sections contain the evaluation of the cask for structural integrity under the hypothetical accident conditions.

2.7.1 Free Drop The RT-100 Cask is required by 10 CFR 71.73(c)(1) [Ref. 2] to demonstrate structural adequacy for a free drop through a distance of 9 meters onto a flat, unyielding, horizontal surface. The cask payload is oriented to strike the surface to inflict the maximum damage. In determining the orientation that produces the maximum damage, the cask is evaluated for impact orientations in which the cask strikes the impact surface on its bottom end and side. Evaluation of each drop orientation is performed by using finite element analysis techniques. A complete description of the 3-D model used to analyze the cask body is presented in Section 2.6.7.2. The results of each drop orientation listed above are presented in this section. The impact limiters are evaluated in Appendix 2.12 for all loading conditions and orientations. These analyses provide the inertial loads (maximum g-loads) imparted to the cask for each drop orientation (Table 2.12.6-1). Cask body decelerations used in NCT and HAC finite element analyses are shown in Table 2.7.1-1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.7.1-1 Deceleration Loadings in RT-100 Cask Body Finite Element Analyses Case End Drop (g) Side Drop (g)

HAC (Drop Height = 9.0 m) 123 226 NCT (Drop Height = 0.3 m) 44 52 The mass of the contents is considered when evaluating impact and environmental temperature for the drop is between -40°C and 38°C. For the accident condition, stresses arising from thermal expansion are not considered. However, for determination of properties, the temperatures are considered. The mean normal operating pressure of 241 (kPa) 35 psig is applied in the finite element models to produce the bounding critical stress condition in conjunction with the other loads previously discussed. A separate analysis evaluates the stresses associated with the accident pressure of 588 kPa (85.3 psig) that results from the regulatory fire event. Note that 85.3 psig accident pressure was employed as an equivalent static pressure load uniformly applied on the interior surface of the cask for HAC finite element structural analyses [Ref. 35, Section 7.2.2], this pressure may appear in psia unit in various sections of the SAR. Closure lid bolt preload is considered (Appendix 2.13 and Section 2.6.7.2.2) and fabrication stresses are discussed (Appendix 2.14). The following method and assumptions are adopted in all the hypothetical accident drop analyses:

The following sections contain the evaluation of the RT-100 for impact orientations in which the cask strikes the impact surface on its bottom end and side. The impact conditions (in accordance with Regulatory Guide 7.8 [Ref. 3] and the categories of load to be considered for the hypothetical accident conditions) are similar to those for the 0.3 meter free drops under normal conditions of transport as discussed in Section 2.6.7. Therefore, the discussions in the following sections refer to Section 2.6.7 wherever applicable.

Three categories of loadclosure lid bolt preload, internal pressure, and inertial body loadsare considered on the cask. The inertia loads imposed upon the cask by the impact limiter result from the mass of the entire assembly being acted upon by a design deceleration value of 123 g for the 30-ft end-drop case. The closure lid bolt preload, internal pressure load, and contents loads considered for the 30-ft end-drop condition are similar to those considered for 1-ft end-drop condition in Section 2.6.7.2, with the exception that thermal stresses are not considered for accident conditions.

The material properties of the components are considered to be temperature dependent.

The allowable stress limits criteria are discussed in Section 2.6.7.1. These criteria are used to determine the allowable stresses for each cask component, conservatively using the maximum operating temperature within a given component to determine the allowable stress throughout that component. For cask body analyses presented in this section, the maximum heat conditions (thermal condition 1) are 38°C ambient temperature, maximum decay heat load, and maximum solar insolation.

During fabrication of the RT-100, thermal stresses can be introduced in the inner and outer shells as a result of pouring molten lead between them. Residual stresses may be induced in the inner shell (containment boundary) and the outer shell due to shrinkage of the lead shielding subsequent to lead pouring operations; however, these stresses are relieved early in the life of the cask because Robatel Technologies, LLC Page 2-69

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 of the low creep strength of lead. Therefore, the effects of stresses resulting from the cask fabrication processes are considered negligible. Further discussion of fabrication stresses is provided in Appendix 2.14.

2.7.1.1 End Drop In accordance with the requirements of 10 CFR 71.73(c)(1) [Ref. 2], the RT-100 is structurally evaluated for the 30-foot end-drop condition. In this hypothetical accident, the cask including the payload, spacer (if appropriate), and the impact limiters falls 30 feet onto a flat, unyielding, horizontal surface. The cask strikes the surface in a vertical position and results in an end impact on the bottom of the cask. The types of loading involved in an end-drop accident are closure lid bolt preload, internal pressure, and inertial body load. Section 2.6.7.2 describes the application of each loading condition.

2.7.1.1.1 End Drop Evaluation In accordance with the requirements of 10 CFR 71.73(c)(1) [Ref. 2], the RT-100 is structurally evaluated for the 30-foot end-drop condition. In this hypothetical accident, the cask including the payload and the impact limiters falls 30 feet onto a flat, unyielding, horizontal surface. The cask strikes the surface in a vertical upright position. For the RT-100 cask, the bottom end drop is bounding. In the bottom down position, the prying load on the closure bolts is maximized.

Stress results for the 9-meter bottom end drop combined are documented in Table 2.7.1-2. The table documents the primary membrane (Pm), primary membrane plus primary bending (Pm+Pb) stresses in accordance with the criteria presented in Regulatory Guide 7.6 [Ref. 4].

As shown in Table 2.7.1-2, the margins of safety when compared to the stress intensity for each category are positive. The most critically stressed component in the system is the flange; this result is due to bending as a result of the inertial loads on the cask lids. The minimum margin of safety is found to be +1.5 for primary membrane plus bending stress intensity. The locations of the critical sections correspond to the maximum stress location shown in Figure 2.7.1-1 through Figure 2.7.1-6.

2.7.1.1.2 Lead Slump Evaluation The following sections provide the lead slump evaluation of the RT-100. During an end drop accident, the shielding capability of the RT-100 cask may be reduced as a result of lead slump.

2.7.1.1.2.1 Elastic Deformation The maximum lead slump occurs during the previously analyzed bottom end drop in Section 2.7.1.1.1. The relative displacement is obtained from the finite element analysis. Figure 2.7.1-7 shows the exaggerated displacement plot under this drop orientation. The total elastic displacement of the lead column is 1.62 mm.

2.7.1.1.2.2 Plastic Deformation with Maximum Gap Maximum plastic deformation of the lead shield occurs when the package experiences extreme cold conditions prior to the end drop. During extreme cold conditions, the contraction of the lead shield forms a small gap at the top of the lead column. The reduced height of the lead shield due to contraction is:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 hlead = hlead (1 + T) = 2037.4 mm

Where, hlead = 2040.9 mm Initial height of lead shield at 21.1°C

= 2.78x10-5 mm/mm/°C Coefficient of thermal expansion for lead at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference The reduced height of the annular column formed by the steel shells due to contraction is:

hsteel = hsteel (1 + T) = 2039.0 mm

Where, hsteel = 2040.9 mm Initial height of annular column at 21.1°C

= 1.48x10-5 mm/mm/°C Coefficient of thermal expansion for steel at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference Radial Thermal Expansion In addition to the gap formed in the axial direction, radial gaps also form during extreme cold conditions. For this evaluation, the interference fit between the cask inner shell and lead shield is ignored because during thermal contraction, the lead applies pressure to the steel inner shell. Since the yield strength of lead is low compared to the steel shell, the lead will conform to the shape of the inner shell. Therefore, the lead volume is not lost during the contraction process and the physical gap between lead and outer shell if any will be significantly less than the values predicted in this calculation. The reduced outside radius of the lead shield at -40°C is:

ro = router (1 + T) = 983.3 mm

Where, router = 985.0 mm Initial outside radius of lead shield = inner radius of steel outer shell at 21.1°C

= 2.78x10-5 mm/mm/°C Coefficient of thermal expansion for lead at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference The change in inside radius of lead shield at -40°C:

ri = rinner (1 + T) = 893.6 mm

Where, rinner = 895.1 mm Inner radius of lead shield at 21.1°C

= 2.78x10-5 mm/mm/°C Coefficient of thermal expansion for lead at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference The reduced inside radius of the outer steel shell at -40°C:

ro = rint (1 + T) = 984.1 mm

Where, rint = 985.0 mm Inner radius of steel outer shell at 21.1°C Robatel Technologies, LLC Page 2-71

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 1.48x10-5 mm/mm/°C Coefficient of thermal expansion for steel at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference The change in outside radius of inner steel shell is at -40°C:

ri = rinner (1 + T) = 894.3 mm

Where, rinner = 895.1 mm Inner radius of lead shield at 21.1°C

= 1.48x10-5 mm/mm/°C Coefficient of thermal expansion for steel at -40°C T = -40°C - 21.1°C = -61.1°C Temperature difference Lead Shield Volume The previous section shows that the relative contraction of materials during extreme cold conditions results in a small gap between the lead shield and outer steel shell. The small gap formed in the radial directions is sufficient to allow the lead shield to slump during an HAC bottom impact.

Following exposure to extreme cold conditions (-40°C), the available volume of the lead column is:

Vf = Af x hc = 1.0784x109 mm³

Where, Af = (ro² - ri²) = 5.293x105 mm² Cross-sectional area of lead shield ro = 983.3 mm Outside radius of lead shield at -40°C ri = 893.6 mm Inner radius of lead shield at -40°C hc = 2037.4 mm Height of lead column at -40°C The cross sectional area of the annulus between the inner and outer shells following exposure to extreme cold conditions (-40°C) is:

Ai = (ro² - ri²) = 5.3013x105 mm²

Where, ro = 984.1 mm Inside radius of steel outer shell at -40°C ri = 894.3 mm Outside radius of steel inner shell -40°C Lead Slump Accounting for the contraction of the steel shells and lead shield the reduced height of the lead column based on the net gap is:

hfinal = Vf / Ai = 2034.2 mm Subtracting the reduced height of the lead column from the height of the annular region and ignoring the elastic deformation, the lead slump is:

hslump = hsteel - hfinal = 2039.0 - 2034.2 = 4.8 mm Robatel Technologies, LLC Page 2-72

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.7.1-2 HAC End Drop Stress Summary ANSYS Results RG 7.6 Component and Stress Margin of Allowable Stress State Location Safety (1)

Stress S1 S2 S3 SINT INNER SHELL MPa MPa MPa MPa MPa Pm 7.5 5.7 -30.9 38.4 331 7.6 Inside 12.8 6.5 -51.3 64.1 496 6.7 Pm + P b Center 7.5 5.7 -30.9 38.4 496 11.9 Outside 8.2 -0.5 -11.2 19.4 496 24.6 OUTER SHELL MPa MPa MPa MPa MPa Pm 10.7 0.1 -22.0 32.8 331 9.1 Inside 7.2 -0.2 -26.3 33.5 496 13.8 Pm + P b Center 10.7 0.1 -22.0 32.8 496 14.2 Outside 14.2 0.5 -17.8 32.0 496 14.5 FLANGE MPa MPa MPa MPa MPa Pm -5.2 -11.9 -19.5 14.3 331 22.2 Inside -5.9 -13.2 -20.2 14.2 496 33.8 Pm + P b Center -5.2 -11.9 -19.5 14.3 496 33.8 Outside 4.7 -14.9 -23.9 28.6 496 16.3 OUTER LID MPa MPa MPa MPa MPa Pm 10.1 -2.3 -30.1 40.3 331 7.2 Inside -29.7 -48.1 -104.5 74.8 496 5.6 Pm + P b Center 10.1 -2.3 -30.1 40.3 496 11.3 Outside 68.5 45.1 24.1 44.4 496 10.2 INNER LID MPa MPa MPa MPa MPa Pm 45.2 31.4 9.3 35.9 331 8.2 Inside 47.0 -14.6 -143.5 190.4 496 1.6 Pm + P b Center 45.2 31.4 9.3 35.9 496 12.8 Outside 172.0 77.5 33.6 138.4 496 2.6 Note: (1) The margin of safety is the ratio of the Allowable Stress and the Stress Intensity (SINT) minus 1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-1 RT-100 HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-74

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-2 RT-100 Inner Shell HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-75

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-3 RT-100 Outer Shell HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-76

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-4 RT-100 Flange HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-77

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-5 RT-100 Outer Lid HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-78

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-6 RT-100 Inner Lid HAC End Drop Stress Intensity Results Robatel Technologies, LLC Page 2-79

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-7 RT-100 Lead Slump Robatel Technologies, LLC Page 2-80

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.1.2 Side Drop In accordance with the requirements of 10 CFR 71.73(c)(1) [Ref. 2], the RT-100 is structurally evaluated for the hypothetical accident 30-foot side drop condition. In this event, the cask including the payload and impact limiters falls 30 feet onto a flat, unyielding, horizontal surface. The package strikes the surface in a horizontal position resulting in a side impact. The types of loading involved in a side drop accident are closure lid bolt preload, internal pressure, and inertial body load.

As previously discussed, stress results for the 9-meter side drop combined loading conditions are documented in Table 2.7.1-3. The table documents the primary membrane (Pm), primary membrane plus primary bending (Pm+Pb), stresses in accordance with the criteria presented in Regulatory Guide 7.6 [Ref. 4].

As shown in Table 2.7.1-3, the margins of safety are positive when compared to the stress intensity for each category. The most critically stressed component in the system is the cask outer shell; this condition is due to ovalization of the cask body and the inertial load of the lead shield. The minimum margin of safety is found to be +0.2 for primary membrane plus bending stress intensity.

The locations of the critical sections correspond to the maximum stress location shown in Figure 2.7.1-8 through Figure 2.7.1-13.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.7.1-3 HAC Side Drop Stress Summary ANSYS Results RG 7.6 Component and Stress Margin of Allowable Stress State Location Safety (1)

S1 S2 S3 SINT Stress INNER SHELL MPa MPa MPa MPa MPa Pm 19.1 -13.7 -140.4 159.6 331 1.1 Inside 20.0 -13.9 -139.7 159.7 496 2.1 Pm + P b Center 19.1 -13.7 -140.4 159.6 496 2.1 Outside 18.3 -13.5 -141.3 159.6 496 2.1 OUTER SHELL MPa MPa MPa MPa MPa Pm -14.2 -129.8 -201.4 187.1 331 0.8 Inside -66.9 -166.2 -472.2 405.3 496 0.2 Pm + P b Center -14.2 -129.8 -201.4 187.1 496 1.7 Outside 73.5 36.5 -95.5 169.0 496 1.9 FLANGE MPa MPa MPa MPa MPa Pm 17.1 -12.5 -145.1 162.2 331 1.0 Inside 16.9 -12.6 -144.6 161.5 496 2.1 Pm + P b Center 17.1 -12.5 -145.1 162.2 496 2.1 Outside 17.3 -12.4 -145.5 162.8 496 2.0 OUTER LID MPa MPa MPa MPa MPa Pm 95.6 0.3 -104.9 200.5 331 0.7 Inside 289.3 35.4 -7.0 296.3 496 0.7 Pm + P b Center 95.6 0.3 -104.9 200.5 496 1.5 Outside -34.4 -94.7 -206.7 172.3 496 1.9 INNER LID MPa MPa MPa MPa MPa Pm -4.3 -14.3 -164.4 160.1 331 1.1 Inside -20.9 -70.1 -371.6 350.6 496 0.4 Pm + P b Center -4.3 -14.3 -164.4 160.1 496 2.1 Outside 64.8 33.1 -1.4 66.3 496 6.5 Note: (1) The margin of safety is the ratio of the Allowable Stress and the Stress Intensity (SINT) minus 1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-8 RT-100 HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-83

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-9 RT-100 Inner Shell HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-84

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-10 RT-100 Outer Shell HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-85

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-11 RT-100 Flange HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-86

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-12 RT-100 Outer Lid HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-87

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.1-13 RT-100 Inner Lid HAC Side Drop Stress Intensity Results Robatel Technologies, LLC Page 2-88

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.1.3 Corner Drop In accordance with the requirements of 10 CFR 71.73(c)(1) [Ref. 2], the RT-100 is structurally evaluated for the hypothetical accident 30-foot corner drop condition. Based on the impact limiter analysis provided in Appendix 2.12, Table 2.7.1-4 demonstrates that the end and side drop accelerations bound the CG over corner drop acceleration.

Table 2.7.1-4 Corner Drop Component Accelerations Side Drop End Drop Corner Drop Corner Drop Axial Corner Drop Lateral Acceleration (g) Acceleration (g) Acceleration (g) Component (g) Component (g) 226 123 116 91.4 71.4 To evaluate the stresses generated in the RT-100 during the corner drop (38° from vertical), the ANSYS [Ref. 28] stress results for the side and end drop evaluations are scaled by the ratio of the end and side drop accelerations and the corner drop axial and lateral component accelerations.

Once scaled, the resulting axial and lateral component stresses are summed and compared to the allowable stress intensity.

Stress results for the 9-meter corner drop combined loading conditions are documented in Table 2.7.1-5. The table documents the primary membrane (Pm), primary membrane plus primary bending (Pm+Pb) stresses in accordance with the criteria presented in Regulatory Guide 7.6

[Ref. 4].

As shown in Table 2.7.1-5, the margins of safety when compared to the stress intensity for each category are positive. The most critically stressed component in the system is the inner lid. The minimum margin of safety is found to be +1.0 for primary membrane plus bending stress intensity.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.7.1-5 HAC Corner Drop Stress Summary RG 7.6 End Drop Side Drop Corner Margin of Stress State Allowable SINT SINT SINT Safety(1)

Stress INNER SHELL MPa MPa MPa MPa Pm 38.4 159.6 79.0 331 3.2 64.1 159.7 98.1 496 4.1 Pm + P b 38.4 159.6 79.0 496 5.3 19.4 159.6 64.9 496 6.7 OUTER SHELL MPa MPa MPa MPa Pm 32.8 187.1 83.5 331 3.0 33.5 405.3 153.0 496 2.2 Pm + P b 32.8 187.1 83.5 496 4.9 32.0 169.0 77.2 496 5.4 FLANGE MPa MPa MPa MPa Pm 14.3 162.2 61.9 331 4.4 14.2 161.5 61.6 496 7.1 Pm + P b 14.3 162.2 61.9 496 7.0 28.6 162.8 72.7 496 5.8 OUTER LID MPa MPa MPa MPa Pm 40.3 200.5 93.3 331 2.5 74.8 296.3 149.2 496 2.3 Pm + P b 40.3 200.5 93.3 496 4.3 44.4 172.3 87.4 496 4.7 INNER LID MPa MPa MPa MPa Pm 35.9 160.1 77.3 331 3.3 190.4 350.6 252.3 496 1.0 Pm + P b 35.9 160.1 77.3 496 5.4 138.4 66.3 123.8 496 3.0 Note: (1) The margin of safety is the ratio of the Allowable Stress and the Stress Intensity (SINT) minus 1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.1.4 Oblique Drops In accordance with the requirements of 10 CFR 71.73(c)(1) [Ref. 2], the RT-100 is structurally evaluated for the hypothetical accident 30-foot oblique drop condition. Based on the following analysis, the cask velocities and stresses generated by an oblique-angle drop are bounded by those produced by the side drop. For a shallow angle drop, it is assumed that no energy is absorbed by the first impact limiter that contacts the impact surface, which causes all of the rotational inertia generated by the cask into the second impact limiter. The analysis is performed according to the following basic inertial equations in Standard Handbook for Mechanical Engineers, 7th Edition

[Ref. 51]

Assumptions:

o The rotational inertia of the cask is approximated by a solid cylinder o The cask does not slide along the impact surface o No gravitational acceleration is assumed to occur after initial contact of the cask with the impact surface The equation for the rotational inertia of a cylinder is:

1 l2 Icyl= Icyl = 4 xMx (r2 + 3 )

Where M = mass of cask r = radius of cask l = length of cask R = distance from CG to corner of impact limiter a = angle of the cask at impact For this configuration, the angular momentum of the cask before impact, L1, is represented by:

l L1= Mx1 x ( - tan (a)) x cos (a) 2 Where 1 = impact velocity After impact the angular momentum, L2, of the cask is:

L2= Iimp x2 2

Where Iimp= Icyl +MxR 2 = angular velocity of cask following impact Substituting the rotational inertia for a cylinder, Icyl:

r2 l2 Iimp= Mx ( + +R2 )

4 12 Because no external moments are applied to the cask, angular momentum is conserved.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Therefore:

L1= L2 Substituting:

l r2 l2 Mx 1 x ( - r x tan (a)) x cos (a) =M x ( + +R2 ) x 2 2 4 12 Solving for the angular velocity, 2, gives:

l

( - r x tan (a))x cos (a) 2= 1 x 2 r2 l2

+ + R2 4 12 The maximum angular velocity occurs when the impact angle equals zero. Therefore, the velocity of the secondary impact is:

s = l x 2 Substituting the angular velocity:

l

( - r x tan (a))x cos (a) s = l x 1 x 2 r2 l2

+ + R2 4 12 The limiting case occurs when the secondary impact velocity equals the initial impact velocity.

Therefore:

s = 1 when the angle a = 0 Solving:

l2 2

1 = r2 l2

+ + R2 4 12 From the figure above:

l2 R2= + r2 4

Substituting and solving:

l2 r2 l2 l2 4 2 5 l2 l

= + + + r2 l = r2 = 7.5 = 2.74 2 4 12 4 12 4 r2 r Therefore:

l

= 1.37 D

Where D = diameter of cask This evaluation shows that cask designs with a length-to-diameter ratio greater than 1.37 may result in oblique impact velocities greater than the side drop. However, the length of the RT-100 Robatel Technologies, LLC Page 2-92

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 is 3316 mm and the diameter is 2587 mm for a length-to-diameter ratio of 1.28. Therefore, impact velocities and resulting stresses in the RT-100 during the oblique drop event are less than those experienced during the side drop.

2.7.1.5 Summary of Results Structural analyses are performed for the RT-100 for hypothetical accident conditions free drop conditions. To evaluate the RT-100, 3D ANSYS [Ref. 28] is used to analyze the governing drop cases. All structural members have a positive margin of safety under worst case loading conditions.

It is concluded that the RT-100 is structurally adequate for the HAC free drop conditions.

Therefore, the requirements of 10 CFR 71.73(c)(1) [Ref. 2] have been satisfied.

2.7.2 Crush In accordance with the requirements of 10 CFR 71.73(c)(2) [Ref. 2], the RT-100 is to be subjected to a dynamic crush test by evaluating the package on essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 500-kg mass from 9 m onto the package. The mass must consist of a solid mild steel plate 1 m x 1 m and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 500 kg, and overall density not greater than 1000 kg/m3 based on external dimension. The crush condition is not applicable since the RT-100 weighs more than 500 kg and overall density is greater than 1000 kg/m3.

2.7.3 Puncture In accordance with the requirements of 10 CFR 71.73(c)(3) [Ref. 2] related to puncture (hypothetical accident condition), the RT-100 Cask is analyzed for structural adequacy (Calculation Package RTL-001-CALC-ST-0403 Rev. 4 [Ref. 36]). The cask is assumed to be in a horizontal position and dropped 1 m onto a 15 cm diameter, mild steel bar, oriented vertically on an unyielding surface. The structural evaluation of the RT-100 is performed by classical elastic analysis and finite element analysis methods.

2.7.3.1 Lid Puncture Finite element analysis methods are used to perform the stress evaluation of the RT-100 for the end puncture conditions. The end puncture is analyzed using a three-dimensional finite element model using the computational modeling software ANSYS [Ref. 28]. To simplify the pin puncture analysis, only the upper end of the cask is considered for this evaluation. Figure 2.7.3-1 shows the pin puncture model.

2.7.3.1.1 Lid Puncture Boundary Conditions The puncture load is applied to a 152 mm (6 in) diameter region which corresponds to a 152 mm diameter pin. The load is simulated with an evenly distributed pressure load equal to the dynamic flow stress of the pin; the dynamic flow stress is taken to be 324 MPa (47,000 psi). As discussed in the cask body analysis, the preload torque is included as an initial condition. In addition, the maximum normal operating pressure of 241 KPa (35 psig) is applied to the interior surface of the RT-100.

2.7.3.1.2 Lid Puncture Results Stress results for the 1-meter pin puncture combined loading conditions are documented in Table Robatel Technologies, LLC Page 2-93

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.3-1. The table documents the primary membrane (Pm), primary membrane plus primary bending (Pm+Pb) stresses in accordance with the criteria presented in Regulatory Guide 7.6 [Ref. 4].

Stresses are linearized across critical sections to determine the membrane and bending stresses and subsequently, are compared with allowable stress intensities.

As shown in Table 2.7.3-1, the margins of safety are positive when compared to the stress intensity for each category. The most critically stressed component in the system is the flange; this condition is due to bending as a result of the pin puncture probe striking the center of the lid. The minimum margin of safety is found to be +0.2 for primary membrane plus bending stress intensity. The locations of critical section correspond to the maximum stress location are shown in Figure 2.7.3-2.

Table 2.7.3-1 HAC Pin Puncture Stress Summary RG 7.6 Margin of Stress State Location S1 S2 S3 SINT Allowable Safety Stress INNER LID MPa MPa MPa MPa MPa Pm -108.6 -109.8 -191.5 82.9 331 3.0 Inside 383.4 382.9 -37.7 421.1 485 0.2 Pm + Pb Center -108.6 -109.8 -191.5 82.9 485 4.9 Outside -342.9 -602.3 -603.3 260.4 485 0.9 Robatel Technologies, LLC Page 2-94

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.3-1 RT-100 ANSYS Puncture Model Figure 2.7.3-2 RT-100 Pin Puncture Stress Intensity Results Robatel Technologies, LLC Page 2-95

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.3.2 Cask Side Puncture The following sections describe the cask side puncture analysis.

2.7.3.2.1 Cask Side Puncture Minimum Wall Thickness A series of pin puncture tests (performed at Oak Ridge National Laboratory) are used to develop an empirical equation to determine the stress in the outer wall of a multi-wall cask as a function of the mass of the cask and the thickness of the cask outer wall material. This equation (Nelm's equation

[Ref. 59]) applies to steel-lead-steel cask wall construction and is used to demonstrate pin puncture adequacy for casks with stainless steel walls; this equation has been the basis for the puncture analysis of several previously licensed casks. Solving Nelms equation [Ref. 59] for the RT-100 outer shell:

W 0.71 t = (S) = 1.16 in (29mm) < 35 mm

where, W = 92,594 lb (42,000 kg), maximum gross weight of the package S = 75,000 psi (517.1 MPa), ultimate tensile strength of the outer shell Nelms equation [Ref. 59] shows that the cask outer shell is sufficient to resist puncture.

2.7.3.2.2 Cask Sidewall Bending Stresses When the cask sidewall impacts the puncture pin, the bending force is:

Mxc b = = 15.3 MPa I

Conservatively assuming the compressive and tensile stresses occur at the same location, the stress intensity is doubled to 30.6 MPa. Therefore, the factor of safety is:

517.1 FS = = 15.7 > 1 30.6

where, Fi xm M = = 1589.2 kN-m, moment due to impact force 4

L m = = 1.16 m, moment arm resulting from impact 2

L = htot - hU - hL = 2.32 m, sidewall length htot = 3312.8 mm, cask total height hU = 498 mm, upper impact limiter height hL = 494 mm, lower impact limiter height Fi = Ks x Ai = 5478.2 kN, impact force Ks = 324 MPa, dynamic flow stress for mild steel (3)

Ai = 4 x d2P = 0.0177 m2, puncture probe area dP = 0.15 m, puncture probe diameter Robatel Technologies, LLC Page 2-96

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Therefore, the RT-100 sidewall successfully resists the regulatory puncture drop.

2.7.3.3 Lead Deformation during Side Puncture Following the postulated side puncture of The RT-100, the cask may experience localized deformation in the outer shell. Behind this localized deformation a slight flattening may occur, and results in shielding loss. To quantify this loss, the local stiffness of the cask wall is determined to calculate the energy absorbed by the package. To calculate the total deformation of the lead shield, it is conservatively assumed that the available potential energy of the 1 meter puncture drop is converted to strain energy.

The maximum deformation occurs during postulated puncture event when the cask strikes the puncture probe approximately mid-span on the cask outer shell. Figure 2.7.3-3 shows the side puncture details. For the purposes of this evaluation, the cask is considered a closed cylinder subjected to a concentrated load at the mid-span. The deformation is obtained from Table 31, Case 9 of Roarks Formulas for Stress and Strain, 6th Edition [Ref. 29]. The deflection of the outer shell due to the applied load is:

P L 0.5 R 1.22 y = [0.48 x (R) x (t) ]

Et where:

L = length of the cylinder R = mean radius of the shell P = applied load E = Youngs modulus Solving for the stiffness P Et k = = L 0.5 R 1.22 y [0.48x( ) x( ) ]

R t The RT-100 is considered a composite cylinder comprised of an outer shell, lead shield, and inner shell. The resulting stiffness of each component is shown below.

2.7.3.3.1 Outer Shell Stiffness 1.989 x1010 x 3.505 x102 k1 = 1.946 0.5 1.003 1.22 = 1.743 x 107 N/m

[0.48x( ) x( 2 ) ]

1.003 3.505 x10 where:

L = 1.946 m R = 1.003 m Robatel Technologies, LLC Page 2-97

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 t = 3.505 x 10-2 m P = 6.972 x 108 N E = 1.989 x 1010 Pa 2.7.3.3.2 Lead Stiffness 1.602 x109 x 8.992 x102 k2 = 0.5 1.22 = 1.191 x 107 N/m 1.946 9.401x101

[0.48x( 1 ) x( 2 ) ]

9.401x10 8.992 x10 where:

L = 1.946 m R = 9.401 x 10-1 m t = 8.992 x 10-2 m P = 1.441 x 108 N E = 1.602 x 109 Pa 2.7.3.3.3 Inner Shell Stiffness 1.989 x1010 x 1.905 x102 k3 = 0.5 1.22 = 4.945 x 106 N/m 1.946 8.801x101

[0.48x( 1 ) x( 2 ) ]

8.801x10 1.905 x10 where:

L = 1.946 m R = 8.801 x 10-1 m t = 1.905 x 10-2 m P = 3.789 x 108 N E = 1.989 x 1010 Pa 2.7.3.3.4 Lead Deformation due to Puncture Load The effective stiffness of the composite section of the cask is:

keff = k1 + k2 + k3 = 3.428 x 107 N/m The energy absorbed during impact is:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 U = 1/2 keff x 2 Assuming the energy absorbed is equal to the total potential energy, the potential energy is calculated as:

P.E. = W x h Setting the energy absorbed during impact equal to the total potential energy the outer shell deformation is:

2(Wxh) 1/2 keff x 2 = Wxh = = 0.050 m k eff where:

W = 42,000 kg H = 1.016 m The deformation of the lead is calculated from the ratio of the effective stiffness and lead stiffness:

k lead = x k 2 = 0.017 m eff Although the deformation is comprised of an elastic and inelastic component, the entire deformation is conservatively assumed to be permanent.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.7.3-3 RT-100 Side Puncture Details Robatel Technologies, LLC Page 2-100

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.4 Thermal For hypothetical accident conditions, the RT-100 cask body provides protection and containment of the contents. Thermal expansion of the bolts is evaluated to ensure the containment boundary is maintained. Similarly, the cask body is evaluated for pressures associated with the fire accident; during the accident, the cask is assumed to be subjected to a fire that produces a surrounding environment of 800C for a period of 30 minutes. The thermal evaluation of the hypothetical fire transient is presented Section in 3.4.

2.7.4.1 Summary of Pressures and Temperatures Cask components temperatures under varying conditions are evaluated using the ANSYS finite element computer code [Ref. 28]. The cask cavity pressure is estimated based on the surface averaged temperature of the inner shell at the cavity side. The detail of the thermal analyses is documented in Chapter 3, Section 3.1. Table 3.1.3-1 presents the normal condition maximum temperature along with the maximum surface averaged temperature of inner shell surface at the cavity side. Chapter 3, Table 3.1.3-2 presents the maximum temperatures under hypothetical accident conditions along with the maximum surface averaged temperature of inner shell surface at the cavity side. The surface averaged temperature of the inner shell at the cavity side is used to predict the gas pressure inside the cask; Chapter 3, Table 3.1.4-1 summarizes the maximum NCT and HAC pressures.

2.7.4.2 Differential Thermal Expansion Per the guidance provided by Regulatory Guide 7.6 [Ref. 4] and ASME Section III Appendix F, stresses resulting from Service Level D (HAC) thermal expansion need not be evaluated. However, differential thermal expansion from exposure to the transient fire event is not of concern for the RT-100 cask components. All the structural materials for the cask body assembly are stainless steel that have the same thermal expansion coefficients, as shown in Table 2.2.1-1.

Proprietary Information Content Withheld Under 10 CFR 2.390(b)

For the RT-100 Cask, the closure bolts are the only components of concern during the fire accident that may experience thermal expansion. The bolting evaluation in Appendix 2.13 evaluates the effects of thermal expansion on the closure bolts.

2.7.4.3 Stress Calculations The following Section evaluates the stresses in the bolts and cask body during hypothetical accident conditions.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.4.3.1 Bolt stresses during fire accident The bolt stress evaluation is presented in Appendix 2.13. The evaluation shows that the bolt stresses are less than the allowables. Therefore, the bolts continue to provide a tight seal and containment is maintained.

2.7.4.3.2 Pressure stress during fire accident In accordance with the requirements of 10 CFR 71.73(c)(4), the RT-100 Cask is structurally evaluated when subjected to an accident internal pressure of 689.4 kPa (100 psia). The pressure is based upon an average cask temperature of 73.1°C. For conservatism, the stress intensity values are compared to allowable stress values at 150°C. To obtain pressure stress results, a uniform internal pressure is applied to the ANSYS finite element model.

2.7.4.4 Comparison with Allowable Stresses The accident pressure stresses are presented in Table 2.7.4-1. The table documents the primary membrane (Pm), primary membrane and plus primary bending (Pm+Pb) stresses in accordance with the criteria presented in Regulatory Guide 7.6. As Table 2.7.4-1 shows, the margins of safety are positive when compared to the stress intensity for each category. The most critically stressed component in the system is the inner lid; this condition is due to prying load at the interface of the closure bolt and lid. The minimum margin of safety is found to be +6.4 for primary membrane plus bending stress intensity. The margins of safety are all positive and thus, the RT-100 satisfies the requirements of 10CFR71.73(c)(4) for thermal HAC.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.7.4-1 HAC Pressure Stress Summary ANSYS Results RG 7.6 Component and Stress Margin of Allowable Stress State Location Safety (1)

S1 S2 S3 SINT Stress INNER SHELL MPa MPa MPa MPa MPa Pm 1.2 0.0 -1.0 2.2 331 Large Inside 1.2 0.0 -1.1 2.3 496 Large Pm + Pb Center 1.2 0.0 -1.0 2.2 496 Large Outside 1.2 0.0 -0.9 2.1 496 Large OUTER SHELL MPa MPa MPa MPa MPa Pm 1.2 0.0 -0.7 1.9 331 Large Inside 1.2 0.0 -0.7 2.0 496 Large Pm + Pb Center 1.2 0.0 -0.7 1.9 496 Large Outside 1.2 0.0 -0.6 1.8 496 Large FLANGE MPa MPa MPa MPa MPa Pm 1.2 0.0 -0.4 1.6 331 Large Inside 1.2 0.0 -0.5 1.7 496 Large Pm + Pb Center 1.2 0.0 -0.4 1.6 496 Large Outside 1.2 0.0 -0.4 1.5 496 Large OUTER LID MPa MPa MPa MPa MPa Pm 1.1 0.1 -0.2 1.3 331 Large Inside 1.1 0.1 -0.3 1.4 496 Large Pm + Pb Center 1.1 0.1 -0.2 1.3 496 Large Outside 1.0 0.1 -0.2 1.2 496 Large INNER LID MPa MPa MPa MPa MPa Pm 0.2 -2.1 -36.5 36.7 331 Large Inside -2.1 -6.2 -64.0 61.9 496 Large Pm + Pb Center 0.2 -2.1 -36.5 36.7 496 Large Outside 4.1 2.0 -10.6 14.7 496 Large Robatel Technologies, LLC Page 2-103

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.7.5 Immersion - Fissile Material This Section is not applicable. The RT-100 does not have any fissile material subject to the requirements of 10 CFR 71.55 [Ref. 2].

2.7.6 Immersion - All Package According to the requirements of 10 CFR 71.73(c)(6) [Ref.2], a package must be subjected to water pressure that is equivalent to: immersion under a head of water of at least 15 meters for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Also, 10 CFR 71.61 [Ref. 2] requires that a packages undamaged containment system be able to withstand an external water pressure of 2000 kPa for a period of not less than one hour without collapse, buckling or in-leakage of water. The outer lid is shown to be structurally adequate for a maximum external dynamic crush pressure of the top impact limiter. Therefore, the RT-100 satisfies all of the immersion requirements for a package that is used for the international shipment of radioactive materials.

2.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2)

This Section is not applicable. The RT-100 is limited to a maximum of 3000 A2.

2.7.8 Summary of Damage The analytical results reported in Section 2.7.1 through 2.7.7 indicate that the damage incurred by the RT-100 during the hypothetical accident is minimal, and such damage does not diminish the cask ability to maintain the containment boundary. A 9-meter drop or a 1-meter pin puncture accident may damage the outer shell and result in a localized reduction in shielding ability.

However, the shielding remains intact to satisfy the accident shielding criteria. Based on the analyses of Section 2.7 through 2.7.7, the RT-100 fulfills the structural and shielding requirements of 10 CFR 71[Ref. 2] for all of the hypothetical accident conditions.

2.8 Accident Conditions for Air Transport of Plutonium This Section is not applicable. The RT-100 cask is not to be used to transport Plutonium by air transport.

2.9 Accident Conditions for Fissile Material Packages for Air Transport This Section is not applicable. The RT-100 is limited by 10 CFR 71 [Ref. 2] for quantities of fissile material. However, the RT-100 is not used to transport any fissile material by air transport.

2.10 Special Form This Section is not applicable. The RT-100 is not to be used to transport special form materials as specified in 10 CFR 71.75 [Ref. 2].

2.11 Fuel Rods This Section is not applicable. The RT-100 is not to be used to transport fuel rods.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13 Appendix - Closure Bolt Evaluation The RT-100 package is designed with two sets of closure bolts: 18 M36 hex head bolts at the secondary lid and 32 M48 hex head bolts at the primary lid. These two sets of bolts are credited with maintaining positive closure of the package under all accident conditions. The purpose of this evaluation is to structurally qualify these bolts for the loadings associated with the normal conditions of transport and the hypothetical accident conditions.

2.13.1 Methodology Bolt loadings under the various normal and accident conditions are determined in accordance with the recommendations of NUREG/CR-6007 [Ref.10]. Stresses resulting from these loads are compared with the design criteria in Section 2.1.2.2. Note that in many cases, calculations are made using exact values, not the rounded numbers shown in intermediate steps. In certain situations, the numbers displayed may not be capable of providing the exact final solution. Using the exact numbers, however, provides the most accurate solution possible. Calculation Package RTL-001-CALC-ST-0203, Rev. 6 [Ref. 60] provides additional information.

2.13.2 Loads The following loads are evaluated in this section:

o Internal pressure loads o Temperature loads o Bolt preload o Impact loads o Puncture loads o External pressure loads o Gasket seating load These loads are combined per NUREG/CR-6007 [Ref. 10] in Section 2.13.3.

2.13.2.1 Internal Pressure Loads Per Table 4.3 of NUREG/CR-6007 [Ref. 10], the forces and moments generated under the internal pressure load are a tensile load Fap, a shear load Fsp, a fixed edge closure force Ffp, and a fixed edge closure moment Mfp. These factors are evaluated for the primary and secondary lid bolts.

2.13.2.1.1 Internal Pressure Loads for Primary Lid Closure Bolts The tensile force per bolt due to internal pressure, Fap, is:

2

( )

=

4

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 1835 mm Nb = Number of Bolts

= 32 Pli = Internal Pressure

= 35 psi = 241.3 kN/m² use 250 kN/m² [Ref. 60]

Plo = External Pressure

= 0 kN/m² (conservative)

Thus, (1.835)2 x (250 0)

= = 20.7 4 x 32 The shear force per bolt due to internal pressure Fsp is:

x x x ( ) x 2

=

2 x x x x (1 )

where, El = Primary Lid Material Elastic Modulus, it means the (SA 240 TYP304/304L)

= 195 GPa at 20° C (Table 2.2.1-1)

Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm Nul = Primary Lid Material Poissons Ratio, (SA 240 TYPE 304/304L)

= 0.31 (Table 2.2.1-2)

Ec = Cask Material Elastic Modulus, (SA 240 TYPE 304/304L)

= 195 GPa at 20° C (Table 2.2.1-1) tl = Primary Lid Thickness

= 210 mm tc = Cask Wall Thickness

= 65 mm (neglecting lead)

The remaining terms are as previously defined. However, this expression for shear force does not apply to the RT-100 cask design because the maximum gap between the lid and cask body (4 mm

= 1741 - 1737) per RT100 PE 1001-1 Rev. H, Detail 1, Chapter 1, Appendix 1.4, Attachment 1.4-

2) is less than the minimum gap between the bolt clearance holes and bolt shank (5.5 mm = 52.5-

47) per RT100 PE 1001-1 Rev. H (Chapter1, Appendix 1.4, Attachment 1.4-2) and Machinerys Handbook [Ref. 27]). Thus, the RT-100 primary lid bolts are not subjected to any shear loads.

Additionally, tolerance stackup based on actual cask dimensions show a minimum clearance of 0.517 mm for the primary lid bolts. Furthermore, thermal expansion and contraction calculations demonstrated that bolt clearance is maintained during NCT and HAC conditions to prevent loading the primary lid bolts in shear.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Therefore, Fsp = 0.0 kN/bolt.

The fixed edge closure force Ffp and moment Mfp are:

D lb (Pli Plo ) 1.920 (250 0)

Ffp = = = 120.0 kN/m 4 4 and, (Pli Plo ) D lb2 (250 0) 1.9202 M fp = = = 28.8 kN-m/m 32 32 2.13.2.1.2 HAC (Fire) Internal Pressure Load for Primary Lid Closure Bolts The HAC (fire) pressure load is calculated for the primary lid bolts using the methods presented in 4.3 in NUREG/CR-6007 [Ref. 10]. The elevated pressure due to fire is shown in Section 3.4.3.2.

x 2 x ( )

=

4 x x ( )

=

4 2 x ( )

=

32

where, Pli = Internal Pressure (HAC Fire)

= 689.4 kN/m² [Section 3.4.3.2.5]

Plo = External Pressure

= 0 kN/m² (conservative)

All other terms were previously defined.

Thus, x 1.8352 x (689.4 0)

=

4 x 32

= 57 kN 1.92 x (689.4 0)

=

4

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 1.922 x (689.4 0)

=

32

= 79.4 kN-m/m 2.13.2.1.3 Internal Pressure Load for Secondary Lid Closure Bolts The secondary lid closure bolt forces and moments are determined using the same methodology as shown for the primary lid bolts (Section 2.13.2.1.1), except that the secondary lid features are incorporated.

The tensile force per bolt due to internal pressure Fas is:

x 2 x ( )

=

4 x

where, Dlg = Outer Seal Diameter

= 850 mm Nb = Number of Bolts

= 18 Pli = Internal Pressure

= 35 psi = 241.3 kN/m² use 250 kN/m² [Ref. 60]

Plo = External Pressure

= 0 kN/m² (conservative)

Thus, x (0.850)2 x (250 0)

=

4 x 18

= 7.9 kN/bolt The maximum gap between the lid and cask body (just 4 mm =748 - 744) is less than the minimum gap between the bolt clearance holes and bolt shank (5.5 mm = 40.5 - 35).

Additionally, tolerance stackup based on actual cask dimensions show a minimum clearance of 0.502 mm for the secondary lid bolts. Furthermore, thermal expansion and contraction calculations demonstrated that bolt clearance is maintained during NCT and HAC conditions to prevent loading the secondary lid bolts in shear. As with the primary lid (Section 2.13.2.1.1),

the shear force per bolt due to internal pressure Fss is:

Fss = 0.0 kN/bolt.

The fixed edge closure force Ffs and moment Mfs are:

Dlb (Pli Plo )

Ffs =

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 M fs =

(Pli Plo ) Dlb2 32 where, Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm All other terms are previously defined. Thus, D lb (Pli Plo ) 0.926 (250 0)

Ffs = = = 57.9 kN/m 4 4 (Pli Plo ) D lb2 (250 0) 0.926 2 and, M fs = = = 6.7 kN-m/m 32 32 2.13.2.1.4 HAC (Fire) Internal Pressure Load for Secondary Lid Closure Bolts The HAC (fire) pressure load is calculated for the secondary lid bolts using the methods presented in 4.3 in NUREG/CR-6007 [Ref. 10]. The elevated pressure due to fire is shown in Section 3.4.3.2.

x 2 x ( )

=

4 x x ( )

=

4 2 x ( )

=

32 All other terms were previously defined.

Thus, x 0.8502 x (689.4 0)

=

4 x 18

= 21.7 kN 0.926 x (689.4 0)

=

4

= 159.6 kN/m 0.9262 x (689.4 0)

=

32

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.2.2 Temperature Loads Temperature differentials and/or differences in the thermal-expansion coefficients of the joint components induce bolt loads. These forces are evaluated Per Table 4.4 of NUREG/CR-6007,

[Ref. 10].

2.13.2.2.1 NCT Temperature Loads for Primary Lid Closure Bolts The temperature load is calculated for elevated (hot) and reduced (cold) temperatures of the lid and the bolt at NCT using the method presented in Table 4.4 in NUREG/CR-6007 [Ref. 10]. The cold case is neglected since the tensile force per bolt is negative for this case, and the hot case is considered in the analysis (+ve). NCT temperatures are also used for the initial conditions of HAC drop. Thus, the tensile force per bolt due to temperature Fat is:

Fathot = 0.25 D b E b (l Tl b Tb )

Fatcold = 0.25 D b E b (l Tlc b Tbc )

where, Db = Nominal diameter of the primary lid bolts

= 48 mm = 0.048 m Eb = Youngs modulus of bolt at 20°C

= 202 GPa (Table 2.2.1-1) l = Thermal expansion coefficient of lid at 100°C

= 16.2x10-6 m/m/°C (Table 2.2.1-1) b = Thermal expansion coefficient of bolt at 20°C

-6

= 11.5 x10 m/m/°C (Table 2.2.1-1)

Tfl = NCT hot temperature of lid, inner shell average

= 71°C (Table 3.1.3-1)

Tfb = NCT hot temperature of bolts

= 70°C (Table 3.1.3-1)

Tflc = NCT cold temperature of lid, inner shell average

= -36°C (Table 3.1.3-1)

Tfbc = NCT cold temperature of bolts

-34.9°C (Table 3.1.3-1)

Ti = Assumed initial temperature of bolts and lid

= 20°C Tl = (Tfl - Ti) = Temperature change (T) of the closure lid (NCT hot)

= 51°C Tb = (Tfb - Ti) = Temperature change of the closure bolt (NCT hot)

= 50°C Tlc = (Tflc - Ti) = Temperature change (T) of the closure lid (NCT cold)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Tbc = (Tfbc - Ti) = Temperature change of the closure bolt (NCT cold)

= -54.9°C

Thus,

= 0.25 x x 0.0482 x 202 x 106 x (16.2 x 106 x 51 11.5 x 106 x 50)

= 91.8 kN/bolt

= 0.25 x x 0.0482 x 202 x 106 x (16.2 x 106 x 56 11.5 x 106 x 54.9)

= -100.8 kN/bolt (neglected)

Shear force per bolt due to temperature Fstp is considered zero because the clamped components (primary lid and cask forged ring) have essentially the same temperature.

2.13.2.2.2 HAC (Fire) Temperature Loads for Primary Lid Closure Bolts The HAC fire temperature load is calculated for the primary lid and bolt using the method presented in Table 4.4 in NUREG/CR-6007 [Ref. 10]. Because there is no significant temperature gradient across thickness of the primary lid, fixed edge moments are neglected, see Figure 3.3.1-8.

Additionally, thermal coefficient properties are conservatively considered at 150°C and 20°C for the lid and bolts, respectively.

Fatfire = 0.25 D b E b (l Tl b Tb )

l = Thermal expansion coefficient of hot lid at 150°C

= 16.6x10-6 m/m/°C (Table 2.2.1-1) b = Thermal expansion coefficient of hot bolt at 20°C

-6

= 11.5x10 m/m/°C (Table 2.2.1-1)

Tflf = HAC fire temperature of lid, inner shell average

= 137.0°C (Table 3.1.3-3)

Tfbf = HAC fire temperature of bolts

= 91.9°C (Table 3.1.3-3)

Ti = Assumed initial temperature of bolts and lid

= 20°C Tl = (Tflf - Ti) = Temperature change (T) of the closure lid (fire)

= 117.0°C Tb = (Tfbf - Ti) = Temperature change of the closure bolt (fire)

= 71.9°C All other terms were previously defined.

= 0.25 x x 0.0482 x 202 x 106 x (16.6 x 106 x 117.0 11.5 x 106 x 71.9)

= 407.7 kN/bolt Robatel Technologies, LLC Page 2-167

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.2.2.3 NCT Temperature Loads for Secondary Lid Closure Bolts The secondary lid closure bolt forces are determined using the same methodology as shown for the primary lid bolts (Section 2.13.2.2.1) and incorporating the secondary lid geometry, material properties and temperatures. Only the bolt diameter in the previous equation must be changed since the primary and secondary lids are constructed of the same materials and experience essentially the same temperature. As shown in the following calculations, the cold case is neglected since the tensile force per bolt is negative, and the hot case is considered in the analysis (+ve). The secondary bolt diameter is 36 mm. The tensile force per bolt due to temperature Fat is:

= 0.25 x x 0.0362 x 202 x 106 x (16.2 x 106 x 51 11.5 x 106 x 50)

= 51.6 kN/bolt

= 0.25 x x 0.0362 x 202 x 106 x (16.2 x 106 x 56 11.5 x 106 x 54.9)

= -55.1 kN/bolt (neglected)

As with the primary lid, the shear force per bolt due to temperature Fsts is considered zero since the clamped components (primary and secondary lids) have essentially the same temperature.

2.13.2.2.4 HAC (Fire) Temperature Loads for Secondary Lid Closure Bolts The HAC fire temperature load is calculated for the secondary lid and bolt using the same method from Section 2.13.2.2.2. The only difference is the bolt diameter which is 36mm for the secondary lid bolts.

= 0.25 x x 0.0362 x 202 x 106 x (16.6 x 106 x 117.0 11.5 x 106 x 71.9)

= 229.3 kN/bolt 2.13.2.3 Bolt Preloads Tightening torques for the primary and secondary lid bolts are respectively, 850 N-m +/-10 % and 350 +/-10 % per Chapter 7, Table 7.4.5-1. The method of analysis is described in Table 4.1 of NUREG/CR-6007 [Ref.10].

2.13.2.3.1 Bolt Preload for Primary Lid Closure Bolts The primary lid bolt preload Fpl is determined as follows (Table 4.1 of NUREG/CR-6007

[Ref. 10]):

T Fpl =

K L x Db Db = Nominal Bolt diameter

= 48 mm K = Nut Factor for empirical relation between applied torque and the achieved preload

= 0.15 (lubricated) minimum (EPRI Good 0.30 (dry) maximum Bolting Practices)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 T = Applied Torque

= 850 N-m +/-10%

To determine maximum preload Fplmax for the primary lid bolts, minimum nut factor, K, of 0.15 (lubricated) and maximum tightening torque of 940 N-m (conservatively bounds the 850 N-m

+10% maximum):

T 940 1kN Fpl = K xD = 0.15x0.048 x 1000N L b

= 130.6 kN The residual torsion moment Mrl is (maximum torque):

M rl = 0.5 Tmax = 0.5 940 = 470 N-m = 0.47 kN-m The residual tensile bolt force Farl is Farl = Fplmax = 130.6 kN The minimum preload Fplmin for the primary lid bolt is calculated by conservatively using a min torque of 760 N-m, and a max nut factor of 0.3.

T 760 1kN Fplmin =K = 0.30 x 0.048 x 1000N = 52.8 kN L xDb where:

T = 760 N-m, Applied torque Db = 0.048 m, Nominal bolt diameter Kl = 0.3, Nut factor The residual torsion moment Mrl is (minimum torque):

M rl = 0.5 Tmin = 0.5 760 = 380 N-m = 0.38 kN-m 2.13.2.3.2 Bolt Preload for Secondary Lid Closure Bolts The maximum secondary lid bolt preload Fpsmax is determined in a manner similar to the primary bolt lids (Section 2.13.2.3.1). Thus, T

Fpl =

K L x Db where, Db = Nominal Bolt diameter

= 36 mm T = Applied Torque

= 350 N-m +/-10%

Other terms are as previously defined. The maximum preload Fpsmax for the secondary lid bolts is obtained by using a nut factor K of 0.15 (lubricated) and a tightening torque of 390 N-m (conservatively bounding the 350 N-m+10% maximum).

Robatel Technologies, LLC Page 2-169

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Tmax 390 Fpsmax = =

K L D b 0.15 36

= 72.2 kN The residual torsion moment Mrs is (maximum torque):

Mrs = 0.5 Tmax = 0.5 390 = 195.0 N-m = 0.20 kN-m The residual tensile bolt force Fars is:

Fars = Fpsmax = 72.2 kN The minimum preload Fpsmin for the secondary lid bolt is calculated by conservatively using a min torque of 310 N-m, and a max nut factor of 0.3.

T 310 1kN Fpsmin = K = 0.30 x 0.036 x 1000N = 28.7 kN L xDb where:

T = 310 N-m, Applied torque Db = 0.036 m, Nominal bolt diameter Kl = 0.3, Nut factor The residual torsion moment Mrs is (minimum torque):

Mrs = 0.5 Tmin = 0.5 310 = 155 N-m = 0.16 kN-m 2.13.2.4 Impact Loads Maximum tension and shear loads in the closure bolts due to the regulatory impact drops are evaluated in accordance with NUREG/CR-6007 [Ref. 10]. Using the NUREG terminology, the primary lid bolts are evaluated as closure bolts for an unprotected lid, and the secondary bolts are evaluated as components of a protected lid. This approach means the primary bolt evaluation includes the impact or inertial forces of the entire cask; the secondary lid bolts are evaluated only for the forces due to the inertia of the secondary lid.

2.13.2.4.1 Dynamic Load Factors Drop impact loadings are generally considered triangular or half-sine loadings; NUREG/CR-3966 [Ref. 17] presents dynamic load factor (DLF) charts for either pulse shape. For this analysis, results are compared and loading with the higher DLF is utilized.

Dynamic load factors for triangular and half sine loadings are shown in Figures 2.3 and Figure 2.15 of NUREG/CR-3966 [Ref. 17]. This information is presented as graphs where the DLF is the ordinate and td/T is the abscissa. The latter quantity td/T is the ratio of the impact duration td and the natural period of the impacting object T.

The period of the lids T is considered for bolt closure analyses. T is determined by the lids lowest mode frequency.

Dynamic Load Factors for Primary Lid Closure Bolts To determine the primary lid frequency, the primary lid and secondary lid are considered a single Robatel Technologies, LLC Page 2-170

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 simply-supported flat circular plate. Thus, (Table 36, Case 11a of Roarks Formulas for Stress and Strain [Ref. 29]):

Resonant Frequency of Primary Lid (with secondary lid attached):

4.99

=

2 4

where, D = Lid Flexural Rigidity E1 t31

=

12(1Nu21 )

gc = conversion factor

= 1000 kg-mm/s2-N w = weight per unit area r = lid bolt radius

= Dlb /2

960 mm RT100 PE 1001-1, Rev. H Appendix 1.4, the primary lid weighs 3648 kg and the secondary lid weighs 857 kg. Thus, (3648+857) w

9602

= 0.001556 kg/mm2 D may be determined from previously defined values:

195 x 103 x 2103

=

12 x (1 0.312 )

= 1.665x1011 N-mm The frequency of the primary lid (with attached secondary lid) is:

1.665 x 1011 x 1000

= 0.7942 x 0.001556 x 9604

= 282 Hz The period of the primary lid is equal to 1/ fl, or T = 1/282 = 0.00354 s. Impact durations for the NCT and HAC impacts range from 0.012 s to 0.045 s (RTL-001-CALC-ST-0401, Rev. 6 [Ref.40]).

Thus, the smallest value of the ratio td/T is 3.389 and the largest is 12.71. With these values, the Robatel Technologies, LLC Page 2-171

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 maximum DLF is determined from Figures 2.3 and 2.15 (NUREG/CR-3966 [Ref.17]) to be less than 1.15. Thus, it is concluded that the DLF for the primary lid bolts may be conservatively bounded by a value of 1.15 for both NCT and HAC drops.

Dynamic Load Factors for Secondary Lid Closure Bolts To determine the secondary lid frequency, the secondary lid is considered a simply-supported flat circular plate. Thus, (Table 36, Case 11a of Roarks Formulas for Stress and Strain [Ref. 29]):

Resonant Frequency of Secondary Lid:

= 0.7942 4

where, D = Lid Flexural Rigidity E1t3 1

=

12(1Nu21 )

gc = conversion factor

= 1000 kg-mm/s2-N w = weight per unit area r = lid bolt radius

= Dlb /2

= 463 mm The secondary lid weighs 857 kg. Furthermore, El = Secondary Lid Material Elastic Modulus, (SA 240 TYPE 304/304L)

= 195 GPa at 20° C (Table 2.2.1-1)

Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm Nul = Secondary Lid Material Poissons Ratio, (SA 240 TYPE 304/304L)

= 0.31 (Table 2.2.1-2) tl = Secondary Lid Thickness

= 110 mm (stainless steel only)

Thus, 857 w=

4632

= 0.001272 kg/mm2 D may be determined from previously defined values:

Robatel Technologies, LLC Page 2-172

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 195 x 103 x 1103

=

12 x (1 0.312 )

= 2.395 x1010 N-mm The frequency of the secondary lid is:

2.395 x 1010 x 1000

= 0.7942 x 0.001272 x 4634

= 508 Hz The period of the secondary lid is equal to 1/ fl, or T = 1/508 = 0.00197 s. Impact durations for the NCT and HAC impacts range from 0.012 s to 0.045 s (RTL-001-CALC-ST-0401, Rev. 6 [Ref.40]).

Thus, the value of td/T is 4.6 or more. With this value, the maximum DLF can be determined from Figures 2.3 and 2.15 (NUREG/CR-6007 [Ref. 10]) to be approaching unity. For consistency with the primary lid bolt analyses, the DLF for the secondary lid bolts is set conservatively to 1.15 for both NCT and HAC drops.

2.13.2.4.2 NCT End Drop Loads The following subsections detail calculations for the NCT end drop load.

2.13.2.4.2.1 Primary Lid Bolts Impact loads in the primary lid bolts due to an end drop are determined using the formulas for evaluating bolt forces/moments generated by impact load applied to an unprotected closure in Table 4.5 of NUREG/CR-6007 [Ref. 10]. An acceleration of 44g is used in this analysis (which is reported in Table 2.7.1-1).

The non-prying tensile bolt force per primary lid bolt Fa is:

1.34 ( ) ( + )

=

where, xi = End Drop Impact Angle

= 90° DLF = 1.15 (Section 2.13.2.4.1) ai = Maximum Impact Acceleration

= 44g (Table 2.7.1-1)

WL = Closure Lid Weight

= 3648 kg (use 3650 kg)

Wc = Cask Payload Weight

= 6804 kg (use 7000kg)

Nb = Number of Bolts

= 32 Robatel Technologies, LLC Page 2-173

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

Thus, 1.34 x sin(90.0) x 1.15 x 44 x (3650 + 7000) x 9.81 1kN Fa = x 32 1000N

= 221.4 kN/bolt As discussed in Section 2.13.2.1.1, the RT-100 primary lid bolts are not subjected to any shear loads. Thus, Fs = 0.0 kN/bolt.

The fixed edge closure lid force Ff is:

1.34 ( ) ( + )

=

where, Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm The remaining terms are as previously defined.

Thus, 1.34 x sin(90.0) x 1.15 x 44 x (3650 + 7000) x 9.81 1kN Ff = x x 1.920 1000N

= 1174.4 kN/m The fixed edge closure lid moment Mf is:

1.34 ( ) ( + )

=

8 All other terms are as previously defined.

Thus, 1.34 (90.0) 1.15 44 (3650 + 7000) 9.81 1

=

8 1000

= 281.9 kN-m/m Refer to Section 2.13.3.1 for the calculations of additional tensile bolt force (Fap_c) caused by the prying action and bolt load combinations.

2.13.2.4.2.2 Secondary Lid Bolts Impact loads in the secondary lid bolts due to an end drop are determined similarly as for the primary lid bolts in Section 2.13.2.4.2.1.

Robatel Technologies, LLC Page 2-174

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The non-prying tensile bolt force per secondary lid bolt Fa is:

1.34 ( ) ( + )

=

where, xi = End Drop Impact Angle

= 90o DLF = 1.15 (Section 2.13.2.4.1) ai = Maximum Impact Acceleration

= 44g (Table 2.7.1-1)

WL = Closure Lid Weight

= 857 kg (use 860 kg)

Wcs = Payload Weight borne by Secondary Lid Nb = Number of Bolts

= 18 Since the payload weight is assumed to be evenly distributed across both the primary and secondary lids, the weight borne by the secondary lid can be obtained by multiplying the payload weight by the ratio of areas, i.e.,

2

= = [ ]

0.785

=( ) 7000 = 1674 kg 3.192

Thus, 1.34 x sin(90.0) x 1.15 x 44 x (860 + 1674) x 9.81 1kN Fa = x 18 1000N

= 93.6 kN/bolt As discussed in Section 2.13.2.1.3, the RT-100 secondary lid bolts are not subjected to any shear loads. Thus, Fs = 0.0 kN/bolt The fixed edge closure lid force Ff is:

1.34 ( ) ( + )

=

where, Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm The remaining terms are as previously defined.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

Thus, 1.34 x sin(90.0) x 1.15 x 44 x (860 + 1674) x 9.81 1Kn Ff = x x 0.926 1000N

= 579.4 /

The fixed edge closure lid moment Mf is:

1.34 ( ) ( + )

=

8 where all terms are as previously defined.

Thus, 1.34 sin(90.0) 1.15 44 (860 + 1674) 9.81 1

=

8 1000

= 67.1 Kn-m/m Refer to Section 2.13.3.2 for the calculations of additional tensile bolt force (Fap_c) caused by the prying action and bolt load combinations.

2.13.2.4.3 HAC End Drop Loads The following subsections detail calculations for the HAC end drop load.

2.13.2.4.3.1 Primary Lid Bolts Impact loads in the primary lid bolts due to an end drop are determined using the formulas for evaluating bolt forces/moments generated by impact load applied to an unprotected closure in Table 4.5 of NUREG/CR-6007 [Ref. 10]. An acceleration of 125g is used in this analysis (which bounds the 123g maximum reported in Section 2.12.4.1). The non-prying tensile bolt force per primary lid bolt Fa is:

1.34 ( ) ( + )

=

where, xi = End Drop Impact Angle

= 90° DLF = 1.15 (Section 2.13.2.4.1) ai = Maximum Impact Acceleration

= 125g (Table 2.7.1-1)

WL = Closure Lid Weight

= 3648 kg (use 3650 kg)

Wc = Cask Payload Weight

= 6804 kg (use 7000kg)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Nb = Number of Bolts

= 32

Thus, 1.34 x sin(90.0) x 1.15 x 125 x (3650 + 7000) x 9.81 1Kn Fa = x 32 1000N

= 628.9 Kn/bolt As discussed in Section 2.13.2.1.1, the RT-100 primary lid bolts are not subjected to any shear loads. Thus, Fs = 0.0 Kn/bolt.

The fixed edge closure lid force Ff is:

1.34 ( ) ( + )

=

where, Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm The remaining terms are as previously defined.

Thus, 1.34 x sin(90.0) x 1.15 x 125 x (3650 + 7000) x 9.81 1kN Ff = x x 1.920 1000N

= 3336.4 kN/m The fixed edge closure lid moment Mf is:

1.34 ( ) ( + )

=

8 All other terms are as previously defined.

Thus, 1.34 (90.0) 1.15 125 (3650 + 7000) 9.81 1

=

8 1000

= 800.7 kN-m/m Refer to Section 2.13.3.3 for the calculations of additional tensile bolt force (Fap_c) caused by the prying action and bolt load combinations.

2.13.2.4.3.2 Secondary Lid Bolts Impact loads in the secondary lid bolts due to an end drop are determined similarly as for the primary lid bolts in Section 2.13.2.4.3.1.

Robatel Technologies, LLC Page 2-177

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The non-prying tensile bolt force per secondary lid bolt Fa is:

1.34 ( ) ( + )

=

where, xi = End Drop Impact Angle

= 90o DLF = 1.15 (Section 2.13.2.4.1) ai = Maximum Impact Acceleration

= 125g (Table 2.7.1-1)

WL = Closure Lid Weight

= 857 kg (use 860 kg)

Wcs = Payload Weight borne by Secondary Lid

= 1674 kg (Section 2.13.2.4.2.2)

Nb = Number of Bolts

= 18

Thus, 1.34 x sin(90.0) x 1.15 x 125 x (860 + 1674) x 9.81 1kN Fa = x 18 1000N

= 266.0 kN/bolt As discussed in Section 2.13.2.1.3, the RT-100 secondary lid bolts are not subjected to any shear loads. Thus, Fs = 0.0 kN/bolt The fixed edge closure lid force Ff is:

1.34 ( ) ( + )

=

where, Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm The remaining terms are as previously defined.

Thus, 1.34 x sin(90.0) x 1.15 x 125 x (860 + 1674) x 9.81 1kN Ff = x x 0.926 1000N

= 1646.1 /

Robatel Technologies, LLC Page 2-178

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The fixed edge closure lid moment Mf is:

1.34 ( ) ( + )

=

8 where all terms are as previously defined.

Thus, 1.34 sin(90.0) 1.15 125 (860 + 1674) 9.81 1

=

8 1000

= 190.5 kN-m/m Refer to Section 2.13.3.4 for the calculations of additional tensile bolt force (Fap_c) caused by the prying action and bolt load combinations.

2.13.2.4.4 Corner Drop Evaluations The closure bolt evaluations for the corner drop impact are conducted very similarly to the end drop analyses in Section 2.13.2.1.3. The cask body acceleration is changed and the impact angle xi is set equal to 52.5° (corresponding to a 37.5° angle between cask axis and vertical line).

Additionally, an acceleration of 120 g is used in this analysis (which bounds the 116 g maximum reported in Section 2.12.4.1). Results are summarized in Table 2.13.2-1.

Table 2.13.2-1 Closure Bolt Loads for 9.0 m Corner-Drop Non-Prying Prying Bending Shear Force, BOLT/LOCATION Tensile Force, Force, Ftp Moment, Mbb Fs (kN/bolt)

Ft (kN/bolt) (kN/m) (kN-m/bolt)

M48x170 Bolts 479.0 265.0 2.0 0.0

/Primary Lid M36x120 Bolts 211.4 147.4 0.8 0.0

/Secondary Lid 2.13.2.4.5 Side Drop Evaluations As shown in Sections 2.13.2.1.1 and 2.13.2.1.3, the gap between the cask body and the primary and second lids is smaller than the gap between the bolts and the bolt clearance holes. Therefore, no shear load is imparted to the bolts from the cask body. Since the side impact drop primarily generates shear loads with respect to the bolts, the primary and secondary closure lid bolts do not receive any significant loading from the side impact drop and are acceptable with respect to the end and corner impact drop.

2.13.2.5 Puncture Loads This section evaluates the results of the various puncture loads.

Robatel Technologies, LLC Page 2-179

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.2.5.1 End Puncture Puncture loads in the primary and secondary closure lid bolts due to a puncture are determined using the formulas for evaluating bolt forces/moments in Table 4.7 of NUREG/CR-6007 [Ref. 10].

2.13.2.5.1.1 Primary Lid Bolts The non-prying tensile bolt force per primary lid bolt Ftp is:

sin(x i ) Pun Ftp =

Nb where, xi = End Drop Impact Angle

= 90o Pun = MIN (Pun1 , Pun2)

Nb = Number of Bolts

= 32 The term Pun is the maximum impact force that can be generated by puncture pin during a normal impact. It is the smaller of:

Pun1 = 0.75 D 2pb S yl Pun2 = 0.6 D pb t l Sul where, Dpb = Puncture bar diameter

= 150 mm (10 CFR 71.73 (c)(3) [Ref. 2])

tl = Closure Lid Thickness

= 110 mm (the secondary lid thickness neglecting the lead)

Syl = Yield Strength of Closure Lid Material (SA 240 304L)

= 172 MPa at 20oC (Table 2.2.1-1)

Sul = Ultimate Strength of Closure Lid Material (SA 240 304L)

= 483 MPa at 20oC (Table 2.2.1-1) thus, Pun1 = 0.75 0.1502 172000

= 9,118.5 kN Pun2 = 0.6 x x 0.150 x 0.110 x 483000

= 15,022 kN Pun = MIN (9118.5, 15022 kN)

= 9,118.5 kN Robatel Technologies, LLC Page 2-180

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 and, sin(90) 9118.5 Ftp =

32

= 285.0 kN/bolt.

As shown in Sections 2.13.2.1.1 and 2.13.2.1.3, the design of the primary and secondary lids prevents shear loads from being applied to the bolts. Thus, Fsp = 0 It is noted that the equation given for Fsp in NUREG/CR-6007 [Ref. 10] also shows Fsp = 0.

The fixed edge closure lid force Ff is:

sin(x i ) Pun Ff =

D lb where, Dlb = Primary Lid Bolt Circle Diameter

1920 mm The remaining terms are as previously defined. Thus, sin(90) 9118.5 Ff

1.920

= 1,511.7 kN/m The fixed edge closure lid moment Mf is:

sin(x i ) Pun Mf =

4

thus, sin(90) 9118.5 Mf =

4

= 725.6 kN-m/m The additional tensile bolt force per bolt Ftp caused by the prying action of the primary lid is (NUREG/CR-6007 Table 2.1 [Ref. 10]):

2 Mf C1 (B Ff ) C2 (B P )

D lb D D lb Ftp = lo Nb C1 + C2

= 517.3 kN/bolt Robatel Technologies, LLC Page 2-181

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

where, P = Bolt Preload per unit Length of Bolt Circle

= 692.9 kN/m (as shown in Section 2.13.2.4.3.1)

B = Non-prying Tensile Bolt Force

= MAX (Ff , P)

C1 = Force Constant

= 1.0 C2 = Second Force Constant 8 E l t 3l (D lo D li ) E lf t 3lf 3 (D D )2 1 N +

lo lb ul D lb

=

Lb bN D 2 b E b

= 3.47 (as shown in Section 2.13.2.4.3.1)

Dlo = Closure Lid Diameter at Outer Edge

= 2016 mm Dli = Closure Lid Diameter at Inner Edge

= 1730 mm tlf = Closure Lid Flange Thickness

= 120 mm Elf = Primary Lid Flange Material Elastic Modulus,

((SA 240 TYPE 304L)

= 195 GPa at 20oC (Table 2.2.1-1)

Lb = Bolt length between the top and bottom surfaces of the closure lid at the bolt circle

= 67 mm The total tension force Fa is:

Fa = Ft + Ftp

= 285.0 + 517.3

= 802.3 kN/bolt The shear force Fs is 0.

The maximum bending moment generated by the applied loads Mbb is:

D lb Kb Mbb = Mf Nb Kb + K1 Robatel Technologies, LLC Page 2-182

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 2.4 kN-m

where, Nb Eb D 4b Kb =

Lb D lb 64

4,167.8 kN E l t 3l K1

D lb 2

3 (1 N ul ) + (1 N ul ) D lb 2 2 lo D

= 234,719 kN 2.13.2.5.1.2 Secondary Lid Bolts The non-prying tensile bolt force per secondary lid bolt Fts is:

sin(x i ) Pun Fts =

Nb

where, xi = End Drop Impact Angle

= 90o Pun = MIN (Pun1 , Pun2)

Nb = Number of Bolts

= 18 Pun was evaluated in Section 2.13.2.5.1.1:

Pun = 9,118.5 kN As shown in Figure 2.7.3-2, the primary and secondary lids act together under the pin puncture load. Therefore, the secondary lid receives only a portion of the impact load from the pin; Pun is reduced by the ratio of the secondary lid volume to the total lid volume.

Vs = Secondary Lid Volume

= Dlb2 t l 4

Vt = Total Lid Volume

= D lbp2 t la 4

where, Dlbp = Closure Lid Bolt Diameter at Primary Lid Bolts

= 1920 mm tlp = Closure Lid Thickness at Primary Lid Bolts Robatel Technologies, LLC Page 2-183

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 210 mm tla = Average Lid Thickness t l + t lp

=

2

thus, 110+ 210 tla =

2

= 160 mm Vs = 0.9262 0.110 4

= 0.067 m3 Vt = 1.9202 0.160 4

= 0.463 m3 Vs Pun = Pun' Vt 0.067 Pun = 9118.5 0.463

1325.6 kN and sin(90) 1325.6 Fts

18

= 73.6 kN/bolt.

As shown in Sections 2.13.2.1.1 and 2.13.2.1.3, the design of the primary and secondary lids prevents shear loads being applied to the bolts. Thus, Fss = 0 It is noted that the equation given for Fss in NUREG/CR-6007 [Ref. 10] also shows Fss = 0.

The fixed edge closure lid force Ff is:

sin(x i ) Pun Ff =

Dlb

= 455.7 kN/m

where, Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm Robatel Technologies, LLC Page 2-184

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The fixed edge closure lid moment Mf is:

sin(x i ) Pun Mf =

4

thus, sin(90) 1325.6 Mf =

4

= 105.5 kN-m/m The additional tensile bolt force per bolt Fts caused by the prying action of the secondary lid is (NUREG/CR-6007 Table 2.1 [Ref. 10]):

2 x

( )

x 1 x ( ) 2 x

= ( )x[ ]

1 + 2

= 164 kN/bolt

where, P = Bolt Preload per unit Length of Bolt Circle

= 446.7 kN/m (as shown in Section 2.13.2.4.3.2)

B = Non-prying Tensile Bolt Force

= MAX(Ff, P)

C1 = Force Constant

= 1.0 C2 = Second Force Constant 8 E l t 3l (D lo D li ) E lf t 3lf 3 (D D )2 1 N +

lo lb ul D lb

=

Lb N b D 2 b E b

= 1.79 (as shown in Section 2.13.2.4.3.2)

Dlo = Closure Lid Diameter at Outer Edge

= 1000 mm Dli = Closure Lid Diameter at Inner Edge

= 745 mm tlf = Closure Lid Flange Thickness

= 80 mm Elf = Secondary Lid Flange Material Elastic Modulus, (SA 240 TYPE 304L)

= 195 GPa at 20oC (Table 2.2.1-1)

Lb = Bolt length between the top and bottom surfaces of the closure lid at the bolt circle Robatel Technologies, LLC Page 2-185

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 43 mm The total tension force Fa is:

Fa = Ft + Fts

= 73.6 + 164

= 237.6 kN/bolt The shear force Fs is 0.

The maximum bending moment generated by the applied loads Mbb is:

D lb Kb Mbb = Mf Nb Kb + K1

0.6 kN/m where, N b E b D4b Kb

b lb L D 64

= 2,396.5 kN (as shown in Section 2.13.2.4.3.2)

E l t 3l K1 =

D lb 2

3 (1 N ul ) + (1 N ul ) D lb 2 2 D lo

= 71,203 kN (as shown in Section 2.13.2.4.3.2) 2.13.2.5.2 Side Puncture In Section 2.13.2.1.1, the gap between the cask body and the primary lid is shown to be smaller than the gap between the M48 bolts and the bolt clearance holes. Therefore, no shear load is imparted to the bolts from the cask body. Further, there are no other loads resulting from side puncture at the bolts. Thus, no significant loads are imparted to the primary and secondary closure lid bolts during a side puncture event.

2.13.2.6 External Pressure Loads in the primary and secondary closure lid bolts due to external pressure are evaluated using the formulas for evaluating bolt forces/moments in Table 4.3 of NUREG/CR-6007 [Ref. 10].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.2.6.1 Primary Lid Bolts The pressure outside the cask Plo in the case of immersion is assumed to be 350 kPa (Calculation Package RTL-001-CALC-TH-0102, Rev. 6 [Ref. 42]). The pressure inside the cask Pli is conservatively taken to be 0 kPa.

The axial force per bolt due to external pressure is:

D lg2 (Pli Plo )

Fa = (Table 4.3 of NUREG/CR-6007[Ref. 10])

4 Nb

where, Dlg = Outside Seal Diameter

= 1835 mm Nb = Number of Bolts

= 32 Since this force is negative (inward acting), the actual resulting bolt force is Fa = 0 since the applied load is supported by the cask wall and not by the bolts.

The fixed edge closure lid force is:

Dlb (Pli Plo )

Ff = (Table of 4.3 NUREG/CR-6007 [Ref. 10])

4

where, Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm

thus, 1.920 (0 350)

Ff =

4

= -168.0 kN/m The fixed edge closure lid moment is:

(Pli Plo ) D lb2 Mf = (Table of 4.3 NUREG/CR-6007 [Ref. 10])

32 (0 350) x 1.9202

=

32

= -40.32 kN-m/m The shear bolt force per bolt is:

E l t l (Pli Plo ) D lb2 Fs = (NUREG/CR-6007 [Ref.10])

2 N b E c t c (1 N ul )

Robatel Technologies, LLC Page 2-187

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= -296.5 kN/bolt The maximum gap between the lid and cask body is less than the minimum gap between the bolt clearance holes and bolt shank (see Section 2.13.2.1.1). Thus, the RT-100 primary lid bolts are not subjected to any shear loads. Therefore, Fs = 0.0 kN/bolt.

2.13.2.6.2 Secondary Lid Bolts The pressure outside the cask Plo in the case of immersion is assumed to be 350 kPa (Calculation Package RTL-001-CALC-TH-0102 Rev. 6 [Ref. 42]). The pressure inside the cask Pli is conservatively taken to be 0 kPa.

The axial force per bolt due to external pressure is:

xD2lg x(Pli Plo )

Fa = 4xNb

= -11.0 kN/bolt Thus, Dlg = Outside Seal Diameter

= 850 mm Nb = Number of Bolts

= 18 Since this force is negative (inward acting), the actual resulting bolt force is Fa = 0 (the load is supported by the cask wall and not by the bolts).

The fixed edge closure lid force is:

D lb (Pli Plo )

Ff =

4

= -81.0 kN/m

where, Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm The fixed edge closure lid moment is:

(Pli Plo ) D lb2 Mf =

32 (0 350) x 0.9262

=

32

= -9.4 kN-m/m Robatel Technologies, LLC Page 2-188

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The shear bolt force per bolt is:

E l t l (Pli Plo ) D lb2 Fs =

2 N b E c t c (1 N ul )

= -64.2 The maximum gap between the lid and cask body is less than the minimum gap between the bolt clearance holes and bolt shank (see Section 2.13.2.1.3). Thus, the RT-100 secondary lid bolts are not subjected to any shear loads. Therefore, Fs = 0.0 kN/bolt.

2.13.2.7 Gasket Seating Load A small closure force is required to maintain a positive seal between the cask lid and the cask body.

However, this closure force is much less than the minimum preloads provided for the closure bolts at the primary and secondary lids. Therefore, the gasket seating load is negligible, and F = 0.

2.13.2.8 Vibration-induced Loads Although vibration-induced loads are not significant, they are considered during normal conditions of transport. The loads that are generated due to vibration are outlined in Table 4.8 of NUREG/CR-6007 [Ref. 10].

2.13.2.8.1 NCT Vibration-induced Loads for Primary Lid Closure Bolts The tensile bolt force per bolt is:

( + )

=

where, VTR = Vibration transmissibility of acceleration between the cask support and the closure lid

= 1 ava = Axial vibration acceleration (2g)

= 19.62 m/s2 Wpl = Primary lid weight

= 3648 kg (use 3650 kg)

Wsl = Secondary lid weight

= 857 kg (use 860 kg)

Nb = Number of bolts

= 32

Thus, 1 x 19.62 x (3650 + 860)

Fa =

32

= 2.8 kN Robatel Technologies, LLC Page 2-189

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The fixed edge force is calculated as follows:

sin() ( + )

=

where, xi = End drop impact angle

= 90° Dlb = Primary lid bolt circle diameter

= 1920 mm all other terms are as previously defined

Thus, Sin(90) x 1 x 19.62 x (3650 + 860)

Ff =

x 1.92

= 14.7 kN/m The fixed edge moment is calculated as follows:

( + )

=

8 all terms were previously defined

Thus, 1 x 19.62 x (3650 + 860)

Mf =

8x

= 3.5 kN-m/m 2.13.2.8.2 NCT Vibration-induced Loads for Secondary Lid Closure Bolts The tensile bolt force per bolt is:

=

where, VTR = Vibration transmissibility of acceleration between the cask support and the closure lid

= 1 ava = Axial vibration acceleration (2g)

= 19.62 m/s2 Wsl = Secondary lid weight

= 857 kg (use 860 kg)

Nb = Number of bolts

= 18 Robatel Technologies, LLC Page 2-190

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

Thus, 1 x 19.62 x 860 Fa =

18

= 0.94 kN The fixed edge force is calculated as follows:

sin()

=

where, xi = End drop impact angle

= 90° Dlb = Secondary lid bolt circle diameter

= 926 mm all other terms are as previously defined

Thus, Sin(90) x 1 x 19.62 x 860 Ff =

x 0.926

= 5.8 kN/m The fixed edge moment is calculated as follows:

=

8 all other terms were previously defined

Thus, 1 x 19.62 x 860 Mf =

8x

= 0.7 kN-m/m 2.13.3 Load Combinations The loadings in Section 2.13.2 are combined to form load cases for the closure bolt analysis per NUREG/CR-6007 [Ref. 10]. The corresponding bolt stresses are obtained and compared to the criteria defined in Section 2.1.2.2. A summary of the loads on the bolts for the primary and secondary lids under the normal conditions of transport and the hypothetical accident conditions is presented in Table 2.13.3-1 and Table 2.13.3-2, respectively.

Robatel Technologies, LLC Page 2-191

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.13.3-1 Primary Lid Bolt Load Summary Non-prying Torsional Fixed-edge Fixed-edge Tensile Force Moment Closure Lid Closure Lid Load Case Applied Load (Fa) (Mt) Force (Ff) Moment (Mf)

(kN/bolt) (kN-m/bolt) (kN/m) (kN-m/m)

Residual Minimum 52.8 0.38 0.0 0.0 Preload Torque Maximum 130.6 0.47 0.0 0.0 Gasket Seating Load 0.0 0.0 0.0 0.0 Internal Pressure 250 kN/m² pressure 20.7 0.0 120.0 28.8 Thermal (NCT) 71°C 91.8 0.0 0.0 0.0 Fire Temperature (HAC) - 407.7 0.0 0.0 0.0 Fire Pressure (HAC) - 57 0.0 330.9 79.4 Drop on 15 cm Puncture 285.0 0.0 1511.7 725.6 diameter pin External Pressure 350 kPa pressure 0.0 0.0 -168.0 -40.3 Free Drop (NCT) Drop from 0.3 m height 221.4 0.0 1174.4 281.9 Free Drop (HAC) Drop from 9 m height 628.9 0.0 3336.4 800.7 Vibration (NCT) 2g 2.8 0.0 14.7 3.5 Table 2.13.3-2 Secondary Lid Bolt Load Summary Non-prying Torsional Fixed-edge Fixed-edge Tensile Force Moment Closure Lid Closure Lid Load Case Applied Load (Fa) (Mt) Force (Ff) Moment (Mf)

(kN/bolt) (kN-m/bolt) (kN/m) (kN-m/m)

Residual Minimum 28.7 0.16 0.0 0.0 Preload Torque Maximum 72.2 0.20 0.0 0.0 Gasket Seating Load 0.0 0.0 0.0 0.0 Internal Pressure 250 kN/m² pressure 7.9 0.0 57.9 6.7 Thermal (NCT) 71°C 51.6 0.0 0.0 0.0 Fire Temperature (HAC) - 229.3 0.0 0.0 0.0 Fire Pressure (HAC) - 21.7 0.0 159.6 18.5 Drop on 15 cm Puncture 73.6 0.0 455.7 105.5 diameter pin External Pressure 350 kPa pressure 0.0 0.0 -81.0 -9.4 Drop from 0.3 m Free Drop (NCT) 93.6 0.0 579.4 67.1 height Free Drop (HAC) Drop from 9 m height 266.0 0.0 1646.1 190.5 Vibration (NCT) 2g 0.94 0.0 5.8 0.7 Robatel Technologies, LLC Page 2-192

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The total tensile bolt force (Fa) is obtained by adding the combined non-prying tensile bolt forces (Fa_c) and the combined prying tensile bolt forces (Fap_c), according to the method presented in Table 4.9-III of NUREG/CR-6007 [Ref. 10].

Fa = Fa_c + Fap_c The non-prying tensile bolt forces are combined per Table 4.9-I of NUREG/CR-6007. The preload and temperature loads are summed together to give Fa_pt, and the remaining loads are added for Fa_al. Because the gasket seating load is zero, the primary contribution to Fa_al is the combination of the internal pressure and free drop. The 0.3 meter and 9-meter drop accelerations are considered for NCT and HAC cases, respectively. Then the maximum of Fa_pt and Fa_al is considered for the combined non-prying tensile bolt load (Fa_c).

Fa_c = Max (Fa_pt and Fa_al) where:

Fa_pt = (operating preload + temperature load)

Fa_al = (impact + pressure + gasket)

The combined prying tensile bolt forces and bending moments are calculated using the combined fixed edge force (Ff_c) and the combined fixed edge moment (Mf_c) per the methods presented in Table 4.9-II, and Table 2.1 and Table 2.2 of NUREG/CR-6007 [Ref. 10]. The NCT impact bounds the NCT vibration loads for the primary and secondary lid bolts, respectively. Therefore, the variables Ff_c and Mf_c are calculated by combining the internal pressure and NCT impact. Ff_c is calculated by adding fixed edge forces of internal pressure and impact. Similarly, Mf_c is calculated by adding fixed edge moments of internal pressure and impact. Because the fixed edge forces and moments from the 9-meter drop bound the pin puncture and external pressure, it is conservatively added with the fixed edge from the internal pressures for the HAC cases.

The sections below detail the NCT and HAC bolt load combinations for the primary and secondary lids and the calculations of Fap_c based on the procedure described in this section.

2.13.3.1 NCT Load Combination for Primary Lid Bolt Below are the combination of non-prying tensile bolt forces based on the procedure in Table 4.9-I of NUREG/CR-6007 [Ref. 10].

Fa_pt = Non-prying tensile forces of NCT loads [Ref. 10]

= Preload + Thermal Load (Table 2.13.3 1)

= 130.6 kN + 91.8 kN = 222.4 kN Fa_al = Non-prying tensile forces of NCT loads (all) [Ref. 10]

= Impact + Pressure + Gasket (Table 2.13.3 1)

= 221.4 kN + 20.7 kN + 0 kN = 242.1 kN Fa_c = Max (Fa_pt and Fa_al) [Ref. 10]

= 242.1 kN Robatel Technologies, LLC Page 2-193

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Below are the combination of prying tensile bolt forces based on the procedure in Table 4.9-II of NUREG/CR-6007 [Ref. 10]. The combined prying tensile bolt forces and bending moments are calculated using the combined fixed edge force (Ff_c) and the combined fixed edge moment (Mf_c) per the methods presented in Table 4.9-II, and Table 2.1 and Table 2.2 of NUREG/CR-6007 [Ref.

10]. Ff_c is calculated by adding fixed edge forces of internal pressure and NCT impact. Similarly, Mf_c is calculated by adding fixed edge moments of internal pressure and NCT impact.

Ff_c = Fixed edge forces of NCT Loads [Ref. 10]

= Internal Pressure + Free Drop NCT (Table 2.13.3-1)

= 120 kN/m + 1174.4 kN/m = 1294.4 kN/m Mf_c = Fixed edge moments of NCT Loads [Ref. 10]

= Internal Pressure + Free Drop NCT (Table 2.13.3-1)

= 28.8 kN-m/m + 281.9 kN-m/m = 310.7 kN-m/m Nb = Number of Bolts

= 32 Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm P = Bolt Preload per unit Length of Bolt Circle [Ref. 10]

32

= a_pt = 222.4 kN x = 1179.9 kN/m lb x 1.92m B = Max (Ff_c and P) [Ref. 10]

= 1294.4 kN/m The additional tensile bolt force per bolt Fap_c caused by the prying action of the primary lid is (Table 2.1 of NUREG/CR-6007 [Ref. 10]):

2 _

1 ( _ ) 2 ( )

_ =( )[ ]

1 + 2

where, C1 = Force Constant

= 1.0 [Ref. 10]

C2 = Second Force Constant (see calculation below)

= 3.47 Dlo = Closure Lid Diameter at Outer Edge Robatel Technologies, LLC Page 2-194

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 2016 mm Other terms are as previously defined in this section.

The Second Force Contant is calculated as follows, 3

8 1 13 ( )

2 = ( ) ( + ) ( )

3 ( ) 2 1 2

where, Dli = Closure Lid Diameter at Inner Edge

= 1730 mm tlf = Closure Lid Flange Thickness

= 120 mm Elf = Primary Lid Flange Material Elastic Modulus, (SA 240 TYPE 304/304L)

= 195 GPa at 20 Lb = Bolt length between the top and bottom surfaces of the closure lid at the bolt circle

= 67 mm Other terms are as previously defined in this Chapter.

Thus, 8 195 x 106 x 0.2103 C2 = ( )x(

3 x (2.016 1.920)2 1 0.31 (2.016 1.730) x 195 x 106 x 0.1203 0.067

+ )x( )

1.920 32 x 0.0482 x 202 x 106 C2 = 3.47

and, Fap_c 2 x 310.7 x 1.920 2.016 1.920 1 x (1294.4 1294.4) 3.47 x (1294.4 1179.9)

=( )[ ]

32 1 + 3.47 Fap_c = 256.2 kN/bolt The total tension force Fa is:

Fa = Fa_c + Fap_c Robatel Technologies, LLC Page 2-195

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Fa = 242.1 kN + 256.2 kN = 498.3 kN The shear force Fs is 0.

The maximum bending bolt moment Mbb is:

= ( ) ( ) _

+ 1

Where, 4

= ( ) ( ) ( )

64 32 202 x 106 0.0484 Kb = ( )x( )x( )

0.067 1.920 64 Kb = 4,167 kN 3

1 =

2 3 [(1 2)

+ (1 )2 ( ) ]

195 x 106 x 0.2103 K1 =

1.920 2 3 x [(1 0.312 ) (1 2

+ 0.31) x (2.016) ] x 1.920 K1 = 234,719 kN

Thus, x 1.920 4,167 Mbb = ( )x( ) x 310.7 32 4,167 + 234,719 Mbb = 1.02 kN-m Average bolt direct stress (Sba), average bolt shear stress (Sbs), maximum bolt bending stress (Sbb), maximum shear stress due to torsional bolt moment (Sbt) and the maximum bolt stress intensity (Sbi) are calculated per Table 5.1 of Ref. 10 as follows.

1.2732 x Fa Sba = ( ) = 338.4 MPa D2 1.2732 x Fs Sbs = ( ) = 0 MPa D2 Robatel Technologies, LLC Page 2-196

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 10.186 x Mbb Sbb = ( ) = 128.2 MPa D3 5.093 x Mt Sbt = ( ) = 31.4 MPa D3 Sbi = ( + )2 + 4 ( + )2 = 470.7 MPa

Where, Fa = 498.3 kN, total tensile load Fs = 0.0 kN, total shear load Mbb = 1.02 kN-m, bending bolt moment Mt = 0.47 kN-m (use 0.5), torsional bolt moment (Table 2.13.3-1)

Db = 0.048 m, bolt nominal diameter p = 0.005 m, bolt thread pitch D = Db - 0.9382 x p = 0.0433 m [Ref. 10]

NCT basic allowable stress limit (Sm):

2

= ( ) x Sy = 544 MPa 3

Where:

Sy = 816 MPa, bolt yield strength (Table 2.2.1-1)

NCT stress ratios are documented according to Table 6.1 of Ref. 10, Rt = Sba / Sm = 0.62, average tensile stress ratio Rs = Sbs / 0.6Sm = 0.0, average shear stress ratio Rt2 + Rs2 = 0.39, tension plus shear < 1 Ri = Sbi / 1.35Sm = 0.64, stress intensity ratio 2.13.3.2 NCT Load Combination for Secondary Lid Bolt Below are the combination of non-prying tensile bolt forces based on the procedure in Table 4.9-I of NUREG/CR-6007 [Ref. 10].

Fa_pt = Non-prying tensile forces of NCT loads [Ref. 10]

= Preload + Thermal Load (Table 2.13.3-2)

= 72.2 kN + 51.6 kN = 123.8 kN Fa_al = Non-prying tensile forces of NCT loads (all) [Ref. 10]

Robatel Technologies, LLC Page 2-197

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= Impact + Pressure + Gasket (Table 2.13.3-2)

= 93.6 kN + 7.9 kN + 0 kN = 101.5 kN Fa_c = Max (Fa_pt and Fa_al) [Ref. 10]

= 123.8 kN Below are the combination of prying tensile bolt forces based on the procedure in Table 4.9-II of NUREG/CR-6007 [Ref. 10]. The combined prying tensile bolt forces and bending moments are calculated using the combined fixed edge force (Ff_c) and the combined fixed edge moment (Mf_c) per the methods presented in Table 4.9-II, and Table 2.1 and Table 2.2 of NUREG/CR-6007 [Ref.

10]. Ff_c is calculated by adding fixed edge forces of internal pressure and NCT impact. Similarly, Mf_c is calculated by adding fixed edge moments of internal pressure and NCT impact.

Ff_c = Fixed edge forces of NCT Loads [Ref. 10]

= Internal Pressure + Free Drop NCT (Table 2.13.3-2)

= 57.9 kN/m + 579.4 kN/m = 637.3 kN/m Mf_c = Fixed edge moments of NCT Loads [Ref. 10]

= Internal Pressure + Free Drop NCT (Table 2.13.3-2)

= 6.7 Kn-m/m + 67.1 Kn-m/m = 73.8 Kn-m/m Nb = Number of Bolts

= 18 Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm P = Bolt Preload per unit Length of Bolt Circle [Ref. 10]

18

= a_pt = 123.8 Kn x x 0.926m = 766.0 Kn/m lb B = Max (Ff_c and P) [Ref. 10]

= 766.0 Kn/m The additional tensile bolt force per bolt Fap_c caused by the prying action of the secondary lid is (Table 2.1 of NUREG/CR-6007 [Ref. 10]):

2 _

1 ( _ ) 2 ( )

_ =( )[ ]

1 + 2

where, C1 = Force Constant Robatel Technologies, LLC Page 2-198

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= 1.0 [Ref. 10]

C2 = Second Force Constant (see calculation below)

= 1.79 Dlo = Closure Lid Diameter at Outer Edge

= 1000 mm Other terms are as previously defined in this section.

The Second Force Contant is calculated as follows, 3

8 1 13 ( )

2 = ( ) ( + ) ( )

3 ( ) 2 1 2

where, Dli = Closure Lid Diameter at Inner Edge

= 745 mm tlf = Closure Lid Flange Thickness

= 80 mm Elf = Secondary Lid Flange Material Elastic Modulus, (SA 240 TYPE 304/304L)

= 195 GPa at 20 Lb = Bolt length between the top and bottom surfaces of the closure lid at the bolt circle

= 43 mm Other terms are as previously defined in this Chapter.

Thus, 8 195 x 106 x 0.1103 C2 = ( ) x (

3 x (1.000 0.926)2 1 0.31 (1.000 0.745) x 195 x 106 x 0.083 0.043

+ )x( )

0.926 18 x 0.0362 x 202 x 106 C2 = 1.79

and, Robatel Technologies, LLC Page 2-199

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2 x 73.8 x 0.926 1.000 0.926 1 x (766 637.3) 1.79 x (766 766)

Fap_c =( )[ ]

18 1 + 1.79 Fap_c = 108.0 kN/bolt The total tension force Fa is:

Fa = Fa_c + Fap_c Fa = 123.8 kN + 108.1 kN = 231.8 kN The shear force Fs is 0.

The maximum bending bolt moment Mbb is:

= ( ) ( ) _

+ 1

Where, 4

= ( ) ( ) ( )

64 18 202 x 106 0.0364 Kb = ( )x( )x( )

0.043 0.926 64 Kb = 2,396 kN 3

1 =

2) 2 2 3 [(1 (1 )

+ ( ) ]

195 x 106 x 0.1103 K1 =

0.926 2 3 x [(1 0.312 ) + (1 0.31)2 x (1.000) ] x 0.926 K1 = 71,203 kN

Thus, x 0.926 2,396 Mbb = ( )x( ) x 73.8 18 2,396 + 71,203 Mbb = 0.39 kN-m Robatel Technologies, LLC Page 2-200

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Average bolt direct stress (Sba), average bolt shear stress (Sbs), maximum bolt bending stress (Sbb), maximum shear stress due to torsional bolt moment (Sbt) and the maximum bolt stress intensity (Sbi) are calculated per Table 5.1 of Ref. 10 as follows.

1.2732 x Fa Sba = ( ) = 284.7 MPa D2 1.2732 x Fs Sbs = ( ) = 0 MPa D2 10.186 x Mbb Sbb = ( ) = 118.4 MPa D3 5.093 x Mt Sbt = ( ) = 30.5 MPa D3 Sbi = ( + )2 + ( + )2 = 407.7 MPa

Where, Fa = 231.8 kN, total tensile load Fs = 0.0 kN, total shear load Mbb = 0.39 kN-m, bending bolt moment Mt = 0.2 kN-m, torsional bolt moment (Table 2.13.3-2)

Db = 0.036 m, bolt nominal diameter p = 0.004 m, bolt thread pitch D = Db - 0.9382 x p = 0.0322 m [Ref. 10]

NCT basic allowable stress limit (Sm):

2

= ( ) x Sy = 544 MPa 3

Where:

Sy = 816 MPa, bolt yield strength (Table 2.2.1-1)

NCT stress ratios are documented according to Table 6.1 of [Ref. 10],

Rt = Sba / Sm = 0.52, average tensile stress ratio Rs = Sbs / 0.6Sm = 0.0, average shear stress ratio Rt2 + Rs2 = 0.27, tension plus shear < 1 Ri = Sbi / 1.35Sm = 0.56, stress intensity ratio Robatel Technologies, LLC Page 2-201

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.3.3 HAC Load Combination for Primary Lid Bolt Below are the combination of non-prying tensile bolt forces based on the procedure in Table 4.9-I of NUREG/CR-6007 [Ref. 10].

Fa_pt = Non-prying tensile forces of HAC loads [Ref. 10]

= Preload + Thermal Load (Table 2.13.3-1)

= 130.6 kN + 91.8 kN = 222.4 kN Fa_al = Non-prying tensile forces of HAC loads (all) [Ref. 10]

= Impact + Pressure + Gasket (Table 2.13.3-1)

= 628.9 kN + 20.7 kN + 0 kN = 649.6 kN Fa_c = Max (Fa_pt and Fa_al) [Ref. 10]

= 649.6 kN Below are the combination of prying tensile bolt forces based on the procedure in Table 4.9-II of NUREG/CR-6007 [Ref. 10]. The combined prying tensile bolt forces and bending moments are calculated using the combined fixed edge force (Ff_c) and the combined fixed edge moment (Mf_c) per the methods presented in Table 4.9-II, and Table 2.1 and Table 2.2 of NUREG/CR-6007 [Ref.

10]. Ff_c is calculated by adding fixed edge forces of internal pressure and HAC impact. Similarly, Mf_c is calculated by adding fixed edge moments of internal pressure and HAC impact.

Ff_c = Fixed edge forces of HAC Loads [Ref. 10]

= Internal Pressure + Free Drop HAC (Table 2.13.3-1)

= 120 kN/m + 3336.4 kN/m = 3456.4 kN/m Mf_c = Fixed edge moments of HAC Loads [Ref. 10]

= Internal Pressure + Free Drop HAC (Table 2.13.3-1)

= 28.8 kN-m/m + 800.7 kN-m/m = 829.5 kN-m/m Nb = Number of Bolts

= 32 Dlb = Primary Lid Bolt Circle Diameter

= 1920 mm P = Bolt Preload per unit Length of Bolt Circle [Ref. 10]

32

= a_pt = 222.4 kN x x 1.92m = 1179.9 kN/m lb B = Max (Ff_c and P) [Ref. 10]

= 3456.4 kN/m The additional tensile bolt force per bolt Fap_c caused by the prying action of the primary lid is (Table 2.1 of NUREG/CR-6007 [Ref. 10]):

Robatel Technologies, LLC Page 2-202

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2 _

1 ( _ ) 2 ( )

_ =( )[ ]

1 + 2

where, C1 = Force Constant

= 1.0 [Ref. 10]

C2 = Second Force Constant (Section 2.13.3.1)

= 3.47 Dlo = Closure Lid Diameter at Outer Edge

= 2016 mm Other terms are as previously defined in this section.

thus, Fap_c 2 x 829.5 x 1.920 2.016 1.920 1 x (3456.4 3456.4) 3.47 x (3456.4 1179.9)

=( )[ ]

32 1 + 3.47 Fap_c = 395.6 kN/bolt The total tension force Fa is:

Fa = Fa_c + Fap_c Fa = 649.6 kN + 395.6 kN = 1045.2 kN The shear force Fs is 0.

Average bolt direct stress (Sba) and average bolt shear stress (Sbs) are calculated per Table 5.1 of Ref. 10 as follows.

1.2732 x Fa Sba = ( ) = 709.8 MPa D2 1.2732 x Fs Sbs = ( ) = 0 MPa D2

Where, Fa = 1045.2 kN, total tensile load Fs = 0.0 kN, total shear load Robatel Technologies, LLC Page 2-203

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Db = 0.048 m, bolt nominal diameter p = 0.005 m, bolt thread pitch D = Db - 0.9382 x p = 0.0433 m [Ref. 10]

HAC allowable tensile stress limit (min (0.7Su, Sy)):

= min (0.7, ) = 721 Where:

Sy = 816 MPa, bolt yield strength (Table 2.2.1-1)

Su = 1030 MPa, bolt ultimate tensile strength (Table 2.2.1-1)

HAC stress ratios are documented according to Table 6.1 of Ref. 10, Rt = Sba / Sm = 0.98, average tensile stress ratio Rs = Sbs / 0.6Sm = 0.0, average shear stress ratio Rt2 + Rs2 = 0.97, tension plus shear < 1 2.13.3.4 HAC Load Combination for Secondary Lid Bolt Below are the combination of non-prying tensile bolt forces based on the procedure in Table 4.9-I of NUREG/CR-6007 [Ref. 10].

Fa_pt = Non-prying tensile forces of HAC loads [Ref. 10]

= Preload + Thermal Load (Table 2.13.3-2)

= 72.2 kN + 51.6 kN = 123.8 kN Fa_al = Non-prying tensile forces of HAC loads (all) [Ref. 10]

= Impact + Pressure + Gasket (Table 2.13.3-2)

= 266.0 kN + 7.9 kN + 0 kN = 273.9 kN Fa_c = Max (Fa_pt and Fa_al) [Ref. 10]

= 273.9 kN Below are the combination of prying tensile bolt forces based on the procedure in Table 4.9-II of NUREG/CR-6007 [Ref. 10]. The combined prying tensile bolt forces and bending moments are calculated using the combined fixed edge force (Ff_c) and the combined fixed edge moment (Mf_c) per the methods presented in Table 4.9-II, and Table 2.1 and Table 2.2 of NUREG/CR-6007 [Ref.

10]. Ff_c is calculated by adding fixed edge forces of internal pressure and HAC impact. Similarly, Mf_c is calculated by adding fixed edge moments of internal pressure and HAC impact.

Ff_c = Fixed edge forces of HAC Loads [Ref. 10]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

= Internal Pressure + Free Drop HAC (Table 2.13.3-2)

= 57.9 kN/m + 1646.1 kN/m = 1704 kN/m Mf_c = Fixed edge moments of HAC Loads [Ref. 10]

= Internal Pressure + Free Drop HAC (Table 2.13.3-2)

= 6.7 kN-m/m + 190.5 kN-m/m = 197.2 kN-m/m Nb = Number of Bolts

= 18 Dlb = Secondary Lid Bolt Circle Diameter

= 926 mm P = Bolt Preload per unit Length of Bolt Circle [Ref. 10]

18

= a_pt = 123.8 kN x x 0.926m = 766.0 kN/m lb B = Max (Ff_c and P) [Ref. 10]

= 1704 kN/m The additional tensile bolt force per bolt Fap_c caused by the prying action of the secondary lid is (Table 2.1 of NUREG/CR-6007 [Ref. 10]):

2 _

1 ( _ ) 2 ( )

_ =( )[ ]

1 + 2

where, C1 = Force Constant

= 1.0 C2 = Second Force Constant (Section 2.13.3.2)

= 1.79 Dlo = Closure Lid Diameter at Outer Edge

= 1000 mm Other terms are as previously defined in this section.

thus, 2 x 197.2 x 0.926 1.000 0.926 1 x (1704 1704) 1.79 x (1704 766)

Fap_c =( )[ ]

18 1 + 1.79 Fap_c = 211.5 kN/bolt Robatel Technologies, LLC Page 2-205

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The total tension force Fa is:

Fa = Fa_c + Fap_c Fa = 273.9 kN + 211.5 kN = 485.4 kN The shear force Fs is 0.

Average bolt direct stress (Sba) and average bolt shear stress (Sbs) are calculated per Table 5.1 of Ref. 10 as follows.

1.2732 x Fa Sba = ( ) = 596.1 MPa D2 1.2732 x Fs Sbs = ( ) = 0 MPa D2

Where, Fa = 485.4 kN, total tensile load Fs = 0.0 kN, total shear load Db = 0.036 m, bolt nominal diameter p = 0.004 m, bolt thread pitch D = Db - 0.9382 x p = 0.0322 m [Ref. 10]

HAC allowable tensile stress limit (min (0.7Su, Sy)):

= min (0.7, ) = 721 Where:

Sy = 816 MPa, bolt yield strength (Table 2.2.1-1)

Su = 1030 MPa, bolt ultimate tensile strength (Table 2.2.1-1)

HAC stress ratios are documented according to Table 6.1 of Ref. 10, Rt = Sba / Sm = 0.83, average tensile stress ratio Rs = Sbs / 0.6Sm = 0.0, average shear stress ratio Rt2 + Rs2 = 0.68, tension plus shear < 1 2.13.4 Fatigue Analysis The fatigue analysis is conducted in accordance with the methods presented in Table 6.2 of NUREG/CR_6007 [Ref. 10] based on the procedures provided in ASME Section III NB-3222.4(e) and NB-3232.3 [Ref. 65], and Appendix I [Ref. 66]. Accordingly, the fatigue analysis is completed for normal conditions of transport using a minimum fatigue strength reduction factor of 4 and Robatel Technologies, LLC Page 2-206

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 ASME fatigue curves I-9.4 [Appendix 2.16] with elastic modulus adjustment per Table 6.2 [Ref.

10].

The fatigue analysis identifies two stress cycles: Normal operation and vibration cycles. Normal operation loads include preload, temperature, and pressure. The vibration loads are documented based on the applied axial vibration acceleration loads. The maximum cumulative usage factor due to alternating stress intensity should be less than one.

2.13.4.1 Fatigue Analysis for Primary Lid Bolts 2.13.4.1.1 Normal Operation Cycles Fatigue due to operating loads is documented based on the maximum NCT cases. The direct stress is obtained by combining the axial and bending stresses. The shear stress is obtained by combining the average shear load and torsional bolt moment. The maximum principal stress (S) is calculated from the combined direct and shear stress.

2 S = 2 +/- (2 ) + ² = 468.7 MPa

where,

= Sba + Sbb = 466.6 MPa (Section 2.13.3.1)

= Sbs + Sbt = 31.4 MPa (Section 2.13.3.1)

The corresponding alternating stresses (Sa) are calculated using a fatigue reduction factor (RF) of 4 [Ref. 66]. Also, since the fatigue curve (ASME Section III, Figure I-9.4) is based on a modulus of elasticity of 30 x 103 ksi, which is different than the modulus of elasticity of the bolt materials, a ratio of the modulus of elasticities is considered in calculating the alternating stress per NB-3232.3 [Ref. 65].

S E Sa = RF ( 2 ) ( Edc ) = 960.1 MPa (139 ksi) a

where, RF = 4, Fatigue strength reduction factor Table 6.2 [Ref. 10]

Edc = 206.9 x 10³ MPa, Modulus of elasticity on design fatigue curve Ea = 202.0 x 10³ MPa, Modulus of elasticity of the bolt material Using the alternating stress calculated and the fatigue curve for a maximum nominal stress 2.7 Sm in ASME Section III, Figure I-9.4 [Appendix 2.16], the corresponding fatigue limits are calculated by interpolating the tabular data given in ASME Section III, Table I-9.0, [Appendix 2.16]. The estimated fatigue limit, N1, for the 960.1 MPa (139 ksi) alternating stress is 527 cycles.

Assuming 500 usage cycles (n1), the usage factor is calculated as follows:

n U1 = N1 = 0.95 1

where, Robatel Technologies, LLC Page 2-207

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 n1 = 500 cycles, assumed lifetime operating cycles N1 = 527 cycles, allowable cycles, Table I-9.0, Figure I-9.4, [Appendix 2.16]

2.13.4.1.2 Vibration Cycles Fatigue due to vibration is documented based on the maximum NCT-vibration loads. The direct stress is obtained from the vibration tensile force. The vibration shear stress is neglected for the protected lid design. The maximum principal stress (S) is calculated from the combined stress.

2 S = +/- (2 ) + ² = 1.9 MPa 2

where, 2.8

= 1.2732 ( )2

= 1.2732 x (0.0433 )2 = 1.9 MPa

= 2.8 kN, Vibration tensile load (Table 2.13.3-1)

= 0.0433m, Diameter used in stress calculations

= 0.0 MPa, Vibration shear stress The corresponding alternating stresses (Sa) are calculated using a fatigue reduction factor (RF) of 4 [Ref. 66]. Also, since the fatigue curve (ASME Section III, Figure I-9.4) is based on a modulus of elasticity of 30 x 103 ksi, which is different than the modulus of elasticity of the bolt materials, a ratio of the modulus of elasticities is considered in calculating the alternating stress per NB-3232.3 [Ref. 65].

S E Sa = RF ( 2 ) ( Edc ) = 3.9 MPa (0.57 ksi) a

where, RF = 4, Fatigue strength reduction factor [Ref. 10]

Edc = 206.9 x 10³ MPa, Modulus of elasticity on design fatigue curve Ea = 202.0 x 10³ MPa, Modulus of elasticity of the bolt material Using the alternating stress calculated and the fatigue curve for a maximum nominal stress 2.7 Sm in ASME Section III, Figure I-9.4 [Appendix 2.16], the corresponding fatigue limits are calculated by interpolating the tabular data given in ASME Section III, Table I-9.0 [Appendix 2.16]. The estimated fatigue limit, N1, for the 3.9 MPa (0.57 ksi) vibration alternating stress is 1.2 x 1015 cycles. An infinite vibration cycle of 1 x 107 cycles is conservatively assumed for vibration cycles. The corresponding usage factor (U1) for normal operating cycles is:

n U1 = N1 = 0.00 1

where, n1 = 1 x 107 cycles, assumed infinite vibration cycles N1 = 1.2 x 1015 cycles, allowable cycles Robatel Technologies, LLC Page 2-208

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.4.2 Fatigue Analysis for Secondary Lid Bolts 2.13.4.2.1 Normal Operation Cycles Fatigue due to operating loads is documented based on the maximum NCT cases. The direct stress is obtained by combining the axial and bending stresses. The shear stress is obtained by combining the average shear load and torsional bolt moment. The maximum principal stress (S) is calculated from the combined direct and shear stress.

2 S = 2 +/- (2 ) + ² = 405.4 MPa (58.8 ksi)

where,

= Sba + Sbb = 403.1 MPa (Section 2.13.3.2)

= Sbs + Sbt = 30.5 MPa (Section 2.13.3.2)

The corresponding alternating stresses (Sa) are calculated using a fatigue reduction factor (RF) of 4 [Ref. 66]. Also, since the fatigue curve (ASME Section III, Figure I-9.4) is based on a modulus of elasticity of 30 x 103 ksi, which is different than the modulus of elasticity of the bolt materials, a ratio of the modulus of elasticities is considered in calculating the alternating stress per NB-3232.3 [Ref. 65].

S E Sa = RF ( 2 ) ( Edc ) = 830.5 MPa a

where, RF = 4, Fatigue strength reduction factor [Ref. 10]

Edc = 206.9 x 10³ MPa, Modulus of elasticity on design fatigue curve Ea = 202.0 x 10³ MPa, Modulus of elasticity of the bolt material Using the alternating stress calculated and the fatigue curve for a maximum nominal stress 2.7 Sm in ASME Section III, Figure I-9.4 [Appendix 2.16], the corresponding fatigue limits are calculated by interpolating the tabular data given in ASME Section III, Table I-9.0, [Appendix 2.16]. The estimated fatigue limit, N1, for the 830.5 MPa (120.5 ksi) alternating stress is 696 cycles.

Assuming 500 usage cycles (n1), the usage factor is calculated as follows:

n U1 = N1 = 0.72 1

where, n1 = 500 cycles, assumed lifetime operating cycles N1 = 696 cycles, allowable cycles, Table I-9.0, Figure I-9.4, [Appendix 2.16]

2.13.4.2.2 Vibration Cycles Fatigue due to vibration is documented based on the maximum NCT-vibration loads. The direct stress is obtained from the vibration tensile force. The vibration shear stress is neglected for the protected lid design. The maximum principal stress (S) is calculated from the combined stress.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2

S = +/- (2 ) + ² = 1.2 MPa 2

where, 0.94

= 1.2732 ( )2

= 1.2732 x (0.0322 )2 = 1.2 MPa

= 0.94 kN, Vibration tensile load (Table 2.13.3-2)

= 0.0322m, Diameter used in stress calculations

= 0.0 MPa, Vibration shear stress The corresponding alternating stresses (Sa) are calculated using a fatigue reduction factor (RF) of 4 [Ref. 66]. Also, since the fatigue curve (ASME Section III, Figure I-9.4) is based on a modulus of elasticity of 30 x 103 ksi, which is different than the modulus of elasticity of the bolt materials, a ratio of the modulus of elasticities is considered in calculating the alternating stress per NB-3232.3 [Ref. 65].

S E Sa = RF ( 2 ) ( Edc ) = 2.5 MPa (0.36 ksi) a

where, RF = 4, Fatigue strength reduction factor [Ref. 10]

Edc = 207 x 10³ MPa, Modulus of elasticity on design fatigue curve Ea = 202 x 10³ MPa, Modulus of elasticity of the bolt material Using the alternating stress calculated and the fatigue curve for a maximum nominal stress 2.7 Sm in ASME Section III, Figure I-9.4 [Appendix 2.16], the corresponding fatigue limits are calculated by interpolating the tabular data given in ASME Section III, Table I-9.0 [Appendix 2.16]. The estimated fatigue limit, N1, for the 2.5 MPa (0.36 ksi) vibration alternating stress is 2.4 x 1016 cycles. An infinite vibration cycle of 1 x 107 cycles is conservatively assumed for vibration cycles. The corresponding usage factor (U1) for normal operating cycles is:

n U1 = N1 = 0.00 1

where, n1 = 1 x 107 cycles, assumed infinite vibration cycles N1 = 2.4 x 1016 cycles, allowable cycles 2.13.5 Seal Integrity The maximum stress analyses in the previous sections are based on criteria for the accident conditions intended to prevent failures by excessive plastic deformation or by the rupture of the bolt. Using the yield stress as the stress limit for average tensile bolt stress, as per NUREG/CR-6007 [Ref. 10], implies that a small amount (0.02%) of plastic deformation is permitted. The following calculations show that the O-rings will continue to provide positive sealing of the closure lids even with this small plastic deformation.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.5.1 Primary Lid Seals The 0.02% bolt plastic deformation permitted in NUREG/CR-6007 [Ref. 10] is distributed over the 67 mm bolt shank dimension shown in Detail 1 of Drawing RT-100 PE 1001-1 Rev. G (Chapter 1, Appendix 1.4, Attachment 1.4-2). This may result in a separation between the primary lid and cask flange mating surfaces of 0.0134 mm (= 67mm x 0.0002). However, the primary lid seals are 12

+/-0.3 mm diameter EPDM rubber and the grooves for these seals are 9.4 +/- 0.15 mm deep (Drawing RT-100 PE 1001-1, Rev. G Appendix 1.4). Thus, the seal is minimally compressed 2.15 mm (= (12 - 0.3) - (9.4 + 0.15)). Considering that EPDM O-rings have a compression set of up to 45% at 150 °C (Figure 2.13.5-1), the minimum compression in the seal is 1.18 mm (= 2.15 -

0.45x2.15). Since the minimum seal compression greatly exceeds the separation due to possible plastic deformation, the primary lid/cask flange containment boundary will remain sealed following an HAC drop event.

2.13.5.2 Secondary Lid Seals The 0.02% bolt plastic deformation permitted in NUREG/CR-6007 [Ref.10] is distributed over the 43 mm bolt shank dimension shown in Detail 2 of Drawing RT-100 PE 1001-1 Rev G, (Chapter 1, Appendix 1.4, Attachment 1.4-2). This may result in a separation between the secondary and primary lid mating surfaces of 0.0086 mm (= 43mm x 0.0002). However, the secondary seals are 12 +/-0.3 mm diameter EPDM rubber and the grooves for these seals are 9.4 +/- 0.15 mm deep (Drawing RT-100 PE 1001-1 Rev G (Chapter 1, Appendix 1.4, Attachment 1.4-2). Thus, the seal is minimally compressed 2.15 mm (= (12 - 0.3) - (9.4 + 0.15)). Considering that EPDM O-rings have a compression set of up to 45% (Figure 2.13.5-1) at 150 °C, the minimum compression in the seal is 1.18 mm (= 2.15 - 0.45x2.15). Since the minimum seal compression greatly exceeds the separation due to possible plastic deformation, the primary to secondary lid containment boundary will remain sealed following an HAC drop event.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.13.5-1 Compression Set vs. Temperature (Figure 2-13 from Parker O-ring Handbook [Ref. 50])

2.13.6 Vent Port Cover Plate O-Ring and Bolt Evaluation The RT-100 cask port cover utilizes a double polymer (EPDM) O-ring configuration face seal to protect the leak test port. For this evaluation the diameter of the outer O-ring is considered to maximize the seating force (Calculation Package RTL-001-CALC-ST-0203, Rev. 6 [Ref.60]). The port cover is sealed with six DIN912 M10 x 30-A4-70 bolts.

2.13.6.1 Vent Port Cover Plate O-Ring Evaluation This section evaluates the vent port cover sealing force and calculates the preload to maintain a tight seal (Calculation Package RTL-001-CALC-ST-0203, Rev. 6 [Ref 60]).

2.13.6.1.1 O-ring Sealing Force The O-ring requires a minimum 3.7 N/mm sealing force (Trelleborg, Appendix 1 [Ref. 58]). The force required to seat the polymer O-ring seal is:

Fs = Yf xC = 3.7x ( x136.6 ) = 1,587.8 N

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Yf = Sealing force C = O-ring circumference 2.13.6.1.2 Vent Port Cover Plate Preload The preload force available to maintain a tight seal that accounts for reduction in preload during HAC is:

PL = Fc - PHAC = 50,522 N

Where, Fc = Available closure force

= Pmin x Nb Pmin = Minimum preload per bolt

= Tmin / k / d Tmin = Minimum torque (-10%, Chapter 7, Table 7.4.5-1)

= 24,300 N-mm k = Nut factor - non-lubricated condition

= 0.3 d = Nominal bolt diameter

= 10 mm Nb = Number of bolts

= 6 PHAC = Loss of preload during HAC [Ref. 10]

= 0.0002 x E x AT

= 8,910 N E = Modulus of elasticity

= 1.89 x 1011 Pa @ 100°C At = Tensile area of the bolt [Ref. 27]

0.97431 2

= 0.7854 d -

n

= 77.6386 mm² n = Number of threads per inch

= 16.93 2.13.6.1.3 Factor of Safety to Maintain a Tight Seal Comparing the available preload force to the load required to maintain a tight seal, the factor of safety is:

50,522 FS = = 31.8 1587.8 2.13.6.2 Bolt Evaluation This section evaluates the vent port cover thread engagement and associate stress.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.13.6.2.1 Thread Engagement For the port cover, the mating internal and external threads are manufactured of materials of equal tensile strengths. To prevent stripping of the external threads, the minimum engagement length, Le, is:

2 At Le =

1 K n,max + 0.57735 n ( E s,min K n,max )

2

= 2.46 mm

Where, Kn,max = 8.676 mm (Machinerys Handbook [Ref. 27])

Es,min = 8.862 mm (Machinerys Handbook [Ref. 27])

The available thread length based on the drawings (RT-100 PE 1001-2 Rev G, Chapter 1, Appendix 1.4, Attachment 1.4-3) is 15.5 mm. Since 15.5 mm > 2.46 mm, there are sufficient threads to prevent stripping of the bolts.

2.13.6.2.2 Thread Shear Evaluation The load necessary to shear the external threads due to the tensile force is:

PS = 0.6 x As x Sy

= 121,044 N

where, 1

As = x n x Le x K n,max [2n + 0.57735 (Es,min K n,max )]

Le = 15.5 mm Sy = 2.06 x 108 Pa @ 100°C The tensile force generated in the bolt is:

T l d2 (d + b)

+ +

2 2cos 4 PB =

= 16,791 N

where, T = Torque

= 29700 N-mm d2 = Min major diameter [Ref. 27]

= d - 3/4H +EI D = 10 mm H = Thread height ignoring flats Robatel Technologies, LLC Page 2-214

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3

= x P = 1.299 mm 2

EI = Fundamental deviation [Ref.27]

= 0.032 P = Thread pitch

= 1.5 mm

= Half thread angle

= 30°

= Coefficient of friction [Ref. 27]

= 0.15 Comparing the load required to shear the external threads with the tensile force generated in the bolt, the factor of safety is:

121,044 FS = 16,791

= 7.2 2.13.6.2.3 Load to Break Bolt The load necessary to break the bolt is:

P = Su x At

= 48,058 N where, Su = 6.19 x 108 Pa @ 100°C Since the load required to break the bolt is less than the applied force (48,058 N > 16,791 N), the bolts will not fail.

Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Proprietary Information Content Withheld Withheld Under Under 10 10 CFRCFR 2.390(b) 2.390(b) 2.15 Appendix - Seal Region Stress Evaluation To provide assurance that the primary and secondary cask seals meet the linear elastic requirements of Regulatory Guide 7.6 [Ref. 4] the contact stresses that represent the maximum nodal stresses on the sealing surfaces and the linearized nodal stresses in the solid elements that comprise the seal regions are evaluated and compared to the yield strength of the material at the maximum NCT temperature. The evaluation shows that the RT-100 seal region does not undergo inelastic deformation.

2.15.1 Seal Region Post-Processing Methodology The cask body calculation reports the primary membrane and membrane plus bending stress intensities averages across linearized sections. To evaluate the stresses in the lid gasket region, Robatel Technologies, LLC Page 2-220

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 the lid component is first selected as shown in Figure 2.15.4-1. As shown in Figure 2.15.4-2, the elements comprising the lid gasket region are then selected to evaluate the stresses specific to the primary gasket location. Visual inspection of the model shows the location of the peak stress in the gasket region and the nodes are identified to calculate the average stress. The ANSYS finite element program [Ref. 28] calculates the average stress across a section by identifying the nodal points using the ANSYS APDL commands PATH, PDEF and PPATH. Figure 2.15.4-2 provides an example of where 2 points are defined across the gasket region. Once the points and path are defined, the ANSYS APDL command PRSECT reports the average stresses.

2.15.2 Stress Concentration Factors Trapezoidal grooves are cut into the primary and secondary lids to allow the gaskets to properly seat during the closure process. Figure 2.15.4-3 shows the lid/gasket geometry. Under load, the grooves can cause a stress riser at the radius, r, where the groove transitions from horizontal to vertical (Standard Handbook for Mechanical Engineers [Ref. 55]). For this evaluation, the load is in the form of a bending moment. Using the dimensions provided in Figure 2.15.4-3 and Table 2.15.4-1, the resulting stress concentration factors are calculated in terms of the ratios of D/d and r/d. For the primary and secondary lids, the stress concentration factors are 2.6 and 2.2, respectively.

2.15.3 Seal Region Stress Results The contact stresses that represent the maximum nodal stresses on the sealing surfaces are summarized in Table 2.15.4-2. The resulting membrane plus bending stresses are compared to the yield stress of the material at the maximum NCT temperature for each HAC case.

For the lid gasket grooves, the linearized stress is calculated for each peak stress location as described in Section 2.15.1 and multiplied by the stress concentration factor calculated in Section 2.15.2. Table 2.15.4-3 provides a summary of the resulting factored stress values. As the table shows the minimum factor of safety is 1.2 in accordance with Regulatory Guide 7.6 [Ref.4], the RT-100 seal region experiences no inelastic deformation during all HAC events.

2.15.4 Displacement Results To determine whether the seal remains tight during HAC, the relative displacement of each sealing surface is determined. Table 2.15.4-4 calculates the relative displacement for each sealing surface and Figure 2.15.4-4 through Figure 2.15.4-9 provide graphical representations of the displacement for each case. From the containment evaluation, the permanent plastic deformation for the EPDM O-ring is approximately 25%. Therefore, based on the seal dimensions, the maximum permissible gap is 1.61mm. Reviewing the relative displacements from Table 2.15.4-4, the maximum separation that occurs is 0.07706mm. Since this maximum separation is less than the permissible gap, the seals are predicted to remain tight during all HAC events.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.15.4-1 Stress Concentration Factors r/d D/d 0.01 0.02 0.04 0.06 0.1 0.15 0.2 0.3 1.01 1.76 1.53 1.37 1.32 1.28 1.25 1.22 1.19 1.02 2.05 1.74 1.52 1.42 1.35 1.28 1.25 1.22 1.05 2.58 2.11 1.77 1.62 1.47 1.40 1.34 1.29 1.10 3.09 2.45 2.00 1.80 1.59 1.49 1.40 1.31 1.20 3.62 2.81 2.23 1.97 1.70 1.55 1.44 1.34 1.50 3.80 2.98 2.38 2.15 1.83 1.63 1.52 1.38 2.00 3.14 2.59 2.23 1.88 1.66 1.54 1.40 3.00 3.30 2.68 2.34 1.93 1.67 1.53 1.38 Table 2.15.4-2 Sealing Surface Stress Summary Primary Seal Secondary Seal Yield Strength Primary Sealing Linearized Secondary Linearized at Max. NCT Surface Contact Stress Intensity Sealing Surface Stress Intensity Accident Seal Temp Stress FS Pm+Pb FS Contact Stress FS Pm+Pb Condition (MPa) (MPa) (MPa) (MPa) (MPa) FS Side Drop 184.2 40.1 4.6 15.6 11.8 5.7 32.3 33.7 5.5 End Drop 184.2 22.8 8.1 14.3 12.9 0.0 N/A 62.7 2.9 Puncture 184.2 93.2 2.0 77.0 2.4 83.4 2.2 89.7 2.1 Table 2.15.4-3 Lid Seal Groove Region Stresses Yield Linearized Linearized Accident Strength at Stress in Stress in Condition Max NCT Stress Primary Lid Maximum Stress Secondary Lid Maximum Seal Temp Concentration Primary Seal Stress Concentration Primary Seal Stress (MPa) (MPa) (MPa) FS (MPa) (MPa) FS Side Drop 184.2 2.6 15.0 38.9 4.7 2.2 66.1 145.3 1.3 End Drop 184.2 2.6 45.1 115.9 1.6 2.2 47.7 102.6 1.8 Puncture 184.2 2.6 59.1 153.6 1.2 2.2 71.8 158.0 1.2 Robatel Technologies, LLC Page 2-222

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 2.15.4-4 HAC Seal Region Displacement Minimum Maximum Location Displacement Displacement (mm) (mm)

HAC Side Drop Primary Lid Sealing Surface -0.047281 -0.33065 Primary Seal Flange Surface -0.043598 -0.40771 Relative Displacement -0.003683 0.07706 Secondary Lid Sealing Surface -0.12481 -0.31864 Secondary Lid Sealing Surface on Primary Lid -0.12238 -0.34072 Relative Displacement -0.00243 0.02208 HAC End Drop Primary Lid Sealing Surface -0.27689 -0.43679 Primary Seal Flange Surface -0.27867 -0.43227 Relative Displacement 0.00178 -0.00452 Secondary Lid Sealing Surface -0.93881 -1.08 Secondary Lid Sealing Surface on Primary Lid -0.93328 -1.1009 Relative Displacement -0.00553 0.0209 Puncture Primary Lid Sealing Surface 0.044821 -0.12329 Primary Seal Flange Surface 0.046104 -0.1212 Relative Displacement -0.001283 -0.00209 Secondary Lid Sealing Surface -0.85594 -1.1537 Secondary Lid Sealing Surface on Primary Lid -0.84898 -1.1475 Relative Displacement -0.00696 -0.0062 Robatel Technologies, LLC Page 2-223

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-1 Stress Intensity Contour Plot of Primary Lid Following End Drop.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-2 Stress Intensity Contour Plot of the Primary Seal Region Robatel Technologies, LLC Page 2-225

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-3 Lid Seal Geometry Robatel Technologies, LLC Page 2-226

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-4 Primary Lid Sealing Surface Displacement during Side drop Robatel Technologies, LLC Page 2-227

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-5 Secondary Lid Sealing Surface Displacement during Side drop Robatel Technologies, LLC Page 2-228

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-6 Primary Lid Sealing Surface Displacement during End drop Robatel Technologies, LLC Page 2-229

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-7 Secondary Lid Sealing Surface Displacement during End drop Robatel Technologies, LLC Page 2-230

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-8 Primary Lid Sealing Surface Displacement during Puncture Robatel Technologies, LLC Page 2-231

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 2.15.4-9 Secondary Lid Sealing Surface Displacement during Puncture Robatel Technologies, LLC Page 2-232

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.16 Appendix - Design Fatigue Curves for High Strength Steel Bolting [Ref. 66]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 2.17 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL, dated March 7, 2012 and the following specific Sections:

71.31(a)(1) 71.31(a)(2) 71.33 71.35(a) 71.71 71.73 71.41(a) 71.45 71.85(b) 71.4 71.73(c)(1) 71.73(c)(3) 71.85 71.43 71.45(a) 71.45(b) 7.51 71.73(b) 71.71(c) 71.55(e) 71.59(a)(2) 71.85(b) 71.55 71.61 71.74 71.55(f) 71.75

3. U.S. Nuclear Regulatory Commission, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, Regulatory Guide 7.8.

4. U.S. Nuclear Regulatory Commission, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, Regulatory Guide 7.6.

5. U.S. Nuclear Regulatory Commission, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 Inches (0.1m), Regulatory Guide 7.11.

6. U.S. Nuclear Regulatory Commission, Fabrication Criteria for Shipping Containers, NUREG/CR-3854, March 1985.

7. ASME Boiler & Pressure Vessel Code 2007 Edition,Section III - Division 1 - Subsection ND, "Class 3 Components", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

8. ASME Boiler & Pressure Vessel Code 2007 Edition,Section III - Division 1 - Subsection NF, "Supports", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

9. U.S. Nuclear Regulatory Commission, SCANS (Shipping Cask Analysis System): A Microcomputer Based Analysis System for Shipping Cask Design Review, NUREG/CR-4554, Volumes 3, 6 and 7, February 1990.

10. U.S. Nuclear Regulatory Commission, Stress Analysis of Closure Bolts for Shipping Casks, NUREG/CR-6007, January 1993.

11. NUREG/CR-0481, "An Assessment of Stress-Strain Data Suitable for Finite-Element Elastic-Plastic Analysis of Shipping Containers," Rack, H. & Knorovsky, G., Sandia Laboratories, Albuquerque, NM, September 1978, Retrieved on August 28, 2013, Retrieved from http://rampac.energy.gov/docs/nrcinfo/NUREG_0481.pdf.

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12. U.S. Nuclear Regulatory Commission, Standard Review Plan for Transportation Packages for Radioactive Material, NUREG-1609, March 31, 1999

13. U.S. Government Code of Federal Regulations, Request for Withholding Information Contained in License Application, 10 CFR 2.790

14. U.S. Nuclear Regulatory Commission, Dynamic Analysis to Establish Normal Shock and Vibration of Radioactive Material Shipping Packages, Volume 3: Final Summary Report, NUREG/CR-2146, Vol. 3, October 1983.

15. U.S. Nuclear Regulatory Commission, Engineering Drawings for 10 CFR Part 71 Package Approvals, NUREG/CR-5502, May 1998.

16. U.S. Nuclear Regulatory Commission, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Wall Thickness Greater than 4 Inches (0.1m), Regulatory Guide 7.12.

17. U.S. Nuclear Regulatory Commission, Methods for Impact Analysis of Shipping Containers, NUREG/CR-3966, November 1987.

18. U.S. Nuclear Regulatory Commission, Puncture Testing of Shipping Packages under 10 CFR Part 71, Bulletin 97-02, September 23, 1997.

19. U.S. Nuclear Regulatory Commission, Recommended Welding Criteria for Use in the Fabrication of Shipping Containers for Radioactive Materials, NUREG/CR-3019, March 1985.

20. U.S. Nuclear Regulatory Commission Bulletin, 97-02.

21. U.S. Nuclear Regulatory Commission, Shock and Vibration Environments for a Large Shipping Container During Truck Transport (Part II), NUREG/CR-0128, August 1978.

22. U.S. Nuclear Regulatory Commission, Dynamic Analysis to Establish Normal Shock and Vibration of Radioactive Material Shipping Packages, NUREG-2146, Volumes 1, 2 and 3, dated January 1, 1981-March 31, 1981; April 1, 1981-June 30, 1981; and October 1993, respectively.

23. U.S. Nuclear Regulatory Commission, International Agreement Report, International Code Assessment and Applications Program: Summary of Code Assessment Studies Concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B, December 1993.

24. Bickford, J. & Looram, M., "Good Bolting Practices - A Reference Manual for Nuclear Power Plant Maintenance Personnel, Volume 1: Large Bolt Manual," Yalesville, CT: Electric Power Research Institute, 1987.

25. Blodgett, O. W., Design of Welded Structures, The James F. Lincoln Arc Welding Foundation, Cleveland, Ohio.

26. AISC, Guide to Design Criteria for Bolted and Riveted Joints, 2nd Edition, 2007.

27. Oberg, Erik, Machinerys Handbook, 26th Edition.

28. ANSYS, Release 14.0, ANSYS, Inc., Canonsburg, PA, October, 2011

29. Young, Warren C., Roarks Formulas for Stress and Strain, 6th Edition.

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30. U.S. Nuclear Regulatory Commission, Shock and Vibration Environments for a Large Shipping Container During Truck Transport NUREG-0128.

31. ASME Boiler & Pressure Vessel Code 2010 Edition,Section II - Part D, "Materials", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

32. ASME Boiler & Pressure Vessel Code 2007 Edition,Section III - Division 1 - Subsection NB, "Class 1 Components", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

33. RTL-001-CALC-ST-0201, Rev. 5, "Lifting Structural Evaluation" (PROPRIETARY)

34. RTL-001-CALC-ST-0202, Rev. 4, "Tie-Down Evaluation" (PROPRIETARY)

35. RTL-001-CALC-ST-0402, Rev. 4, "Cask Body Structural Evaluation" (PROPRIETARY)

36. RTL-001-CALC-ST-0403, Rev. 4, "Pin Puncture Evaluation" (PROPRIETARY)

37. WM2001 Conference paper, Benchmarking of LS-DYNA for Use with Impact Limiters, Joseph C. Nichols III, Michael E. Cohen, Robert A. Johnson, 2001.

38. RTL-001-CALC-TH-0102, Rev. 6, "RT-100 Cask Maximum Normal Operating Pressure Calculation" (PROPRIETARY)

39. Bickford, J. & Looram, M., "Good Bolting Practices - A Reference Manual for Nuclear Power Plant Maintenance Personnel, Volume 1: Large Bolt Manual," Yalesville, CT: Electric Power Research Institute, 1987.

40. RTL-001-CALC-ST-0401, Rev. 6, "RT-100 Cask Impact Limiter Drop Evaluation" (PROPRIETARY)

41. U.S. Nuclear Regulatory Commission, Methods for Impact Analysis of Shipping Containers, NUREG/CR-3966.

42. RTL-001-CALC-TH-0102, Rev. 6, "RT-100 Cask Maximum Normal Operating Pressure Calculation" (PROPRIETARY)

43. RTL-001-CALC-TH-0202, Rev. 6, "RT-100 Cask Hypothetical Accident Condition Maximum Pressure Calculation" (PROPRIETARY)

44. ASME B1.13M-2005, METRIC SCREW THREADS: M PROFILE.

45. PAP 008, Specification D'approvisionnement - Mousse Polyurethane - Emballage de TRANSPORT RT-100 (Procurement Specification - Polyurethane Foam - Packaging of TRANSPORT RT-100), Rev. D, ROBATEL Industries (PROPRIETARY)

46. RES 001, Safety Analysis Robatel Package Model RT-100 Drop Test Report, Rev. E, ROBATEL Industries (PROPRIETARY)

47. Drawing 102885 MD 2021-06 Rev. D, "Robatel Transport Package RT100" (PROPRIETARY)

48. Certificate of Conformance for Purchase Order #117039 (Certificate for RT100 Scaled Foam Model) dated 09-07-2012 (PROPRIETARY)

49. U.S. Nuclear Regulatory Commission, Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material, Regulatory Guide 7.9.

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50. Parker O-Ring Handbook ORD 5700, Retrieved on August 28, 2013, Retrieved from http://www.parker.com/literature/ORD%205700%20Parker_O-Ring_Handbook.pdf.

51. Baumeister T. and Marks, L.S. "Standard Handbook for Mechanical Engineers, 7th Edition".

New York: McGraw-Hill Book Co., 1967.

52. U.S. Nuclear Regulatory Commission Bulletin, 96-04.

53. U.S. Nuclear Regulatory Commission Interim Staff Guidance, Use of Computational Modeling Software, ISG-21.

54. Glenn Lee, Radiation Resistance of Elastomers, IEEE Transactions on Nuclear Science, Vol.

NS-32, No. 5, October 1985.

55. Baumeister T. and Marks, L.S. "Standard Handbook for Mechanical Engineers, 9th Edition".

New York : McGraw-Hill Book Co., 1987.

56. ANSI N14.6-1978, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10000 pounds (4500 kg) or More for Nuclear Materials," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

57. KTA 3905, Load Attaching Points on Loads in Nuclear Power Plants, Safety Standards of the (German) Nuclear Safety Standards Commission, June 1999 Edition including rectification of July 2000.

58. TRELLEBORG Sealing Solutions O-Ring and Backup Rings Catalog, August 2011 Edition

59. Shappert, L.B. The Radioactive Materials Packaging Handbook. Oak Ridge, Tennessee:

Oak Ridge National Laboratory, 1988. ORNL/M-5003.

60. RTL-001-CALC-ST-0203, Rev. 6, RT-100 Bolting Calculation (PROPRIETARY)

61. GENERAL PLASTICS Design Guide for LAST-A-FOAM FR-3700 Crash & Fire Protection of Radioactive Material Shipping Containers, Rev. 02.20.12

62. J. F. Harvey, Theory and Design of Pressure Vessels, New York: Van Nostrand Reinhold, 1991.

63. CN-13039-203, Rev. 0, RT-100 Cask Lead Shrinkage Evaluation (PROPRIETARY)

64. CN-21004-201, Rev. 1, RT-100 Cask Bolting Load Combination Verification (PROPRIETARY)

65. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code (BPVC): Section III - Rules for Construction of Nuclear Facility Components - Division 1 -

Subsection NB - Class 1 Components, 2010.

66. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code (BPVC): Section III - Rules For Construction of Nuclear Facility Components," Division 1-Appendices, 2010.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

3. THERMAL EVALUATION Robatel has performed a thermal evaluation of the RT-100 using the Nuclear Industry standards and under the RT Company Quality Assurance Program [Ref. 1]. This thermal evaluation shows that the RT-100 meets or exceeds all the 10 CFR 71 regulatory requirements [Ref. 2]. The thermal review is based in part on the descriptions and evaluations presented in the General Information Chapter 1 and Structural Evaluation Chapter 2 of the application. Similarly, results of the thermal review are considered in the review of several other sections of the application. An example of information flow for the thermal review is shown in Figure 3-1.

RT identified, described, discussed, and analyzed the principal thermal engineering design of the RT-100, components, and systems that are important to safety. Section 3 describes how the package complies with the performance requirements of 10 CFR 71 [Ref. 2]. Results of the thermal evaluation verified that the thermal performance of the RT-100 design (for both NCT and HAC) meets the thermal regulatory requirements as follows:

o The RT-100 design is evaluated to demonstrate that it satisfies the thermal requirements of 10 CFR 71.31(a)(1) ]; 10 CFR 71.31(a)(2); 10 CFR 71.33, and 10 CFR 71.35(a) [all Ref. 2].

o The application identifies the established codes and standards used for the thermal design according to 10 CFR 71.31(c) [Ref. 2].

o The performance of the RT-100 is evaluated under the tests specified in 10 CFR Part 71.71 [Ref. 2] for NCT and 10 CFR Part 71.73 [Ref. 2] for HAC and also referenced 10 CFR 71.41(a) [Ref. 2].

o The RT-100 is designed, constructed, and prepared for transport so that there is no significant decrease in packaging effectiveness under the tests specified in 10 CFR 71.71 (NCT) and references in 10 CFR 71.43(f) and 71.51(a)(1) [all Ref. 2].

o The RT-100 is designed, constructed, and prepared for transport so that the accessible surface temperature does not exceed the regulatory limits specified in 10 CFR 71.43(g)

[Ref. 2].

o The RT-100 design does not rely on mechanical cooling systems to meet containment requirements in reference to 10 CFR 71.51(c) [Ref. 2].

o The RT-100 has adequate thermal performance to meet the containment, shielding, sub-criticality, and temperature requirements of 10 CFR 71 [Ref. 2] for (NCT/HAC).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3-1 Information Flow for the Thermal Review General Structural Information Evaluation

Dimensions

Deformation

Materials

Displacement of Contents

Decay Heat Load Thermal Review Loading

Decay Heat

Ambient Conditions

Fire

Insolation Evaluation

Material Properties

Modeling

Heat Transfer

Gas Generation

Pressure Analysis - Conduction

- Convection

- Radiation Results

Temperatures

Pressures

Degradation

Gas Inventory

Phase Changes Structural Containment Shielding Criticality Evaluation Evaluation Evaluation Evaluation

Temperatures

Temperatures

Combustion

Not Applicable

Pressures

Pressures

Decomposition

Gas Inventory

Dehydration

Melting

Displacement of Shielding Operating Acceptance Tests Procedures and Maintenance

Temperatures

Thermal Tests

Pressure

Temperatures

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.1 Description of Thermal Design The thermal design aspects of the RT-100 are related primarily to protecting the sensitive components of the cask and the contents from the elevated temperatures produced by the hypothetical fire accident. The primary thermal criteria that are applied to the thermal evaluation are maintaining the lead shielding in the cask body and secondary lid below the melting temperature of lead, and the maximum temperature of the O-ring seals below their maximum operating temperature. The components primarily responsible for maintaining the temperatures of these components below their acceptance criteria are the impact limiters covering the top and bottom of the cask, and the thermal shield on the radial cask surface.

The impact limiters are made from a polyurethane foam material that has a low thermal conductivity. The impact limiters cover the top and bottom ends of the cask. They protect the lead in the bottom of the cask body and the O-rings in the primary lid, secondary lid, and the vent port cover plate. The impact limiters are designed to remain attached to the cask during normal operations and hypothetical accident conditions, and to insulate the lead and O-rings from the high temperatures of the hypothetical fire accident. The thermal shield covering the radial cask surface is made of a ceramic fiber material with a very low thermal conductivity. The ceramic fiber is covered by a thin, stainless steel cover that protects it from damage during normal handling. The ceramic fiber material is designed for use in insulating refractory furnaces, and providing an excellent thermal barrier for the fire accident, thus preventing the radial lead from exceeding its melting point.

The RT-100 is designed to accommodate contents with a maximum decay heat of 200 watts. This low decay heat value does not produce a significant temperature gradient through the cask body, and as a result, no specific design features are required to facilitate removing the heat from the cavity.

3.1.1 Design Features As briefly described in Section 3.1, the RT-100 design has two primary thermal design features:

the impact limiters and the radial thermal shield. These features are identified in Chapter 1, Figure 1.2.1-1 which highlights the primary components of the cask.

3.1.1.1 RT-100 Description The RT-100 cask body consists of inner and outer shells constructed of 304/304L stainless steel.

Lead shielding is provided between these radial shells, as well as between the 304/304L stainless steel bottom forging and bottom plate. The upper end of the cask comprises the upper 304/304L stainless steel forging that is attached to the inner and outer shells, and contains the mating surface for the primary lid. The primary and secondary lids are constructed of 304L stainless steel, as is the cover plate. The primary lid is attached using thirty-two (32) M48 hex head bolts and the secondary lid is secured using eighteen (18) M36 hex head bolts. The upper and lower impact limiter cover each end of the cask, and are constructed of 304L stainless steel shells containing polyurethane foam blocks. The impact limiters are secured to the body via twelve (12) M24 threaded studs. The RT-100 is described in greater detail in Chapter 1, Section 1.2.1.

Most of the outer shell of the cask is covered by a ceramic fiber thermal shield that is secured by a thin 304L stainless steel cover. Other portions of the radial cask surface are covered by the 318 Robatel Technologies, LLC Page 3-3

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 stainless steel tie-down arms and tie-down arm baseplate, and by the 304L stainless steel lifting blocks. The exposed surfaces of the tie-down arm baseplate are covered by the ceramic fiber thermal shield and 304L stainless steel cover.

3.1.1.2 RT-100 Dimensions The RT-100 thermal analysis is performed using the basic cask dimensions as presented in Appendix 1.4. The inner and outer shell thicknesses at the side of the cask are reduced to account for minimum thickness due to manufacturing tolerances. Specifically, the following material thicknesses are reduced by 2mm:

o Thicknesses of the cask body inner and outer stainless steel shells o Bottom end of the cask body o Stainless steel bottom forging welded to the inner shell o Stainless steel bottom plate welded to the outer shell.

To represent the condition of undersized shells with a poured lead fill, lead thickness in both the sidewalls and the cask bottom end are simultaneously increased by 4 mm. This approach is conservative as it reduces the amount of stainless steel material protecting the lead from the HAC fire temperatures and thus, maximizes temperatures in the lead.

In order to maximize the amount of heat that can enter the cask body during the fire, no gaps are assumed between the lead and the outer shell. The lead and stainless steel outer shell are assumed to be in perfect contact. No other gaps are assumed in the thermal evaluation between the various components of the cask.

3.1.2 Contents Decay Heat The RT-100 is designed for a maximum decay heat of 200 watts. This value is selected as the design basis.

The analysis of the cask for normal condition of transport (NCT) and hypothetical accident conditions (HAC) is performed using the ANSYS finite element computer code [Ref. 3]. In this analysis, the decay heat of the contents is modeled as a uniform heat flux on the internal surfaces of the cask cavity.

To calculate this uniform heat flux over the inside surface of the cask, the inside diameter and the height of the cask cavity is used.

Ain = dL + 2(d2/4) where Ain = inside surface area of the cask (m2) d = inside diameter of the cask (m)

L = height of the cavity (m)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Based on the RT-100 drawing provided in Appendix 1.4, the cask inside diameter is 1.730 m and the height of the cavity is 1.956 m. The area is then Ain = x1.730x1.956 + 2(x1.7302/4) = 15.332 m2 The uniform internal heat flux, qint, is then qint = 200W / 15.332 m2 = 13.04 W/m2 3.1.3 Summary Tables of Temperatures Section 3.1.3 presents summary tables of maximum temperatures occurring in the RT-100 as a result of the NCT and HAC evaluations described in detail in Sections 3.3 and 3.4. Limiting temperatures for consideration in the structural and containment evaluations are the maximum temperatures. Therefore, the following tables present maximum temperatures that occur in the various cask components under NCT and HAC. Table 3.1.3-1 presents the NCT maximum temperatures while Table 3.1.3-2 and Table 3.1.3-3 present the maximum temperatures HAC. For the fire accident evaluation, the time at which the component reaches its maximum temperature is listed along with the temperature. In some cases, temperatures are after cessation of the fire transient.

The tables also present the maximum averaged surface temperature of the inner shell at the cavity side. These averaged surface temperatures are used to predict the cavity pressure under normal and hypothetical conditions, respectively.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.1.3-1 RT-100 Maximum Normal Condition Temperature Summary Hot Hot Cold Cold Allowable Component Case 1 Case 2 Case 1 Case 2 Temperature Reference

(°C) (°C) (°C) (°C) (°C)

Primary Seal 68.7 42.1 -35.5 -24.5 150 Ref. 8 Secondary Seal 70.3 42.9 -34.7 -23.8 150 Ref. 8 Quick Disc. Valve Cover Seal 72.5(a) (a) (a) (a) 150 Ref. 8 Lead Shield 73.2 43.1 -34.5 -23.6 328 Ref. 5 (p. 907)

Closure Bolts 70.0 42.8 -34.9 -23.9 (b)

Outer Surface 93.0 41.3 -36.3 -25.4 50/85 10 CFR 71.43(g)

Inner Shell Maximum 73.1 42.6 -35.1 -24.1 Inner Shell Average 71.0 41.7 -36.0 -25.0 Total Impact Limiter Average 67.4 39.5 -38.3 -27.3 Top Impact Limiter Average 72.5 39.5 -38.3 -27.4

a. The NCT maximum temperature of the components surrounding the cover plate is the upper impact limiter average temperature (reported in Table 3.1.3-1) where the temperatures are higher on the external surfaces of the impact limiter. Thus the maximum temperature of the cover plate containment O-ring is considered to be 72.5°C with no further analysis. Since Hot Case 1 is the bounding upper temperature of this O-ring, the other NCT cases are not considered.

b. 10 CFR 71.43(g)A package must be designed, constructed, and prepared for transport so that in still air at 38°C (100°F) and in the shade, no accessible surface of a package would have a temperature exceeding 50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in an exclusive use shipment.

Table 3.1.3-2 RT-100 Maximum Calculated Temperature of Cask under HAC with Pin Puncture Damage on Top Impact Limiter Time After Start Allowable Temperature Component of Fire Temperature Reference

(°C)

(Minutes) (°C)

Primary Seal Maximum 110.8 291.6 150 Ref. 8 Secondary Seal Maximum 131.1 33.4 150 Ref. 8 (c) (c)

Quick Disc. Valve Cover Seal 133.1 33.4 150 Ref. 8 Lead Shield Maximum 304.8 34.5 328 Ref. 5 (p. 907)

Closure Bolts Maximum 133.1 33.4 Cask Body Maximum 799.1 30.0 Inner Shell Average 136.3

c. The port cover plate location is on the primary lid, close to the primary lid closure bolts. The cover plate is thermally insulated by the upper impact limiter. The highest temperature reported on the primary and secondary lid are the closure bolts (where the puncture bar penetrates the impact limiter) with a maximum temperature of 133.1°C (reported in Table 3.1.3-2). This temperature of 133.1°C being bonding to all the lids and cover plate recorded temperatures, the maximum temperature of the cover plate containment O-ring during HAC is considered to be 133.1°C with no further analysis.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.1.3-3 RT-100 Maximum Calculated Temperature of Cask under HAC with Pin Puncture Damage at the Side of the Cask Body Time After Start Allowable Temperature Component of Fire Temperature Reference

(°C)

(minutes) (°C)

Primary Seal Maximum 110.3 285.1 150 Ref. 8 Secondary Seal Maximum 91.3 1624.4 150 Ref. 8 Quick Disc. Valve Cover Seal (d)

(d) 150 Ref. 8 Lead Shield Maximum 304.7 34.5 328 Ref. 5 (p. 907)

Closure Bolts Maximum 91.9 1302.5 Cask Body Maximum 799.1 30.0 Inner Shell Average 137.0

d. The Quick-Disconnect Valve Cover Plate Seal maximum temperature is considered bounded by the result of the top impact limiter pin puncture HAC. Thus, the side puncture result is not reported.

3.1.4 Summary Tables of Maximum Pressures The maximum internal pressures in the RT-100 are determined using the maximum temperatures presented in Table 3.1.3-1, Table 3.1.3-2, and Table 3.1.3-3 above. Details of these pressure calculations are presented in Section 3.3.2 for NCT and in Section 3.4.3 for HAC. Table 3.1.4-1 presents a summary of the maximum pressure calculations for normal and accident conditions.

These pressures are utilized in the structural evaluation presented for the cask body in Sections 2.6 and 2.7.

Table 3.1.4-1 RT-100 Summary of Maximum Normal and Hypothetical Accident Condition Pressures Condition Maximum Pressure Normal Conditions of Transport 342.7 kPa (49.7 psia)

(MNOP)

Hypothetical Accident Conditions 689.4 kPa (100 psia) 3.2 Material Properties and Component Specifications The material properties and specifications for the RT-100 materials of construction are presented in this section. The determination of material properties are carefully evaluated to ensure that for each thermal analysis:

o The appropriate thermal properties for the package materials are correctly incorporated into the thermal evaluations.

o Appropriate expressions are used for conductive, convective, and radiative heat transfer among package components, and from the surfaces of the package to the environment.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.2.1 Material Properties The thermal evaluation of the RT-100 is performed using material properties taken from standard industry references or manufacturer provided data in Tables 3.2.1-1 through 3.2.1-4. The thermal absorptivities and emissivities are appropriate for the package surface conditions and each thermal condition. When reporting a property as a single value, the evaluation shows that this value bounds the equivalent temperature-dependent property. This section includes references for the data provided.

Only room temperature values of conductivity, density, and specific heat are available for General Plastics FR-3700 series LAST-A-FOAM [Ref. 10, 11, and 12]. Quantitative temperature dependent material properties are not provided. However, most of the foam remains at temperatures close to ambient due to the dimensions of the RT-100 impact limiters which result in long heat conduction paths (see Figure 3.3.1-3). Thus, reduction in the foam thermal properties due to elevated temperatures will not be significant. Therefore, the use of temperature-independent thermal properties is justified.

Information on the EPDM O-ring material is provided in TRELLEBORG, Aug. 2011 Edition

[Ref. 8] and PARKER O-RING Handbook [Ref. 16] for two different suppliers. The temperature range specified in Table 3.2.1-1 is conservative from the values specified in those two references. Additional information on the O-rings is presented in Appendices Attachment 3.5-1 and Attachment 3.5-2.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.2.1-1 Temperature-Independent Material Properties Material Properties Reference Page Value Number Density Ref.24: Page 744 8030 kg/m3 Emissivity Ref. 2 0.9 (fire)

Emissivity Stainless Steel 304 Ref. 2 0.8 (cool-down)

Emissivity Ref. 5: Page 750 (normal 0.2 and 929 condition)

Density Ref. 5: Page 907 11340 kg/m3 Lead Melting Point Ref. 5: Page 907 328°C (601 K)

Density 176.2 kg/m3 (11 lb/ft3) 1172.5 J/kg-K @1366.5K Ceramic Paper Ref. 9 Specific Heat (0.28 BTU/lb-°F @

2000°F)

Proprietary Information Content Withheld Under 10 CFR 2.390(b)

Seal (EPDM) Working Ref. 8 & 16 -45°C to 150°C Temperature Robatel Technologies, LLC Page 3-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.2.1-2 Temperature-Dependent Material PropertiesStainless Steel 304

[Ref. 13, page 765]

Specific Heat Thermal Conductivity Temperature (°C)

(J/kg-K) (W/m-K) 20 472.6 14.8 50 483.6 15.3 75 493.1 15.8 100 499.4 16.2 125 506.7 16.6 150 511.4 17.0 175 520.1 17.5 200 525.7 17.9 225 530.0 18.3 250 532.5 18.6 275 536.5 19.0 300 541.7 19.4 325 545.5 19.8 350 547.7 20.1 375 551.4 20.5 400 552.3 20.8 425 557.0 21.2 450 557.8 21.5 475 562.3 21.9 500 563.1 22.2 525 566.3 22.6 550 568.1 22.9 575 571.2 23.3 600 572.9 23.6 625 575.9 24.0 650 577.5 24.3 675 580.4 24.7 700 581.9 25.0 725 585.8 25.4 750 587.2 25.7 Robatel Technologies, LLC Page 3-10

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.2.1-3 Temperature-dependent Material PropertiesLead

[Ref. 5, page 907]

Specific Heat Thermal Conductivity (W/m- K)

Temperature (°C)

(J/kg- K)

-173.15 118 3.97E+01

-73.15 125 3.67E+01 26.85 129 3.53E+01 126.8 132 3.40E+01 5

326.8 142 3.14E+01 5

Table 3.2.1-4 Temperature-dependent Material PropertiesCeramic Paper

[Ref. 9]

Thermal Conductivity Temperature (°C)

(W/m-K) 93.3 4.759E-02 204.4 5.206E-02 315.6 5.912E-02 426.7 6.907E-02 537.8 8.219E-02 648.9 9.834E-02 760.0 1.174E-01 871.1 1.396E-01 3.2.2 Component Specifications This section includes the technical specifications of RT-100 components that are important to the thermal performance, as illustrated by the following examples:

o In the case of seals, the operation temperature limits o Maximum allowable service temperatures for package components o Minimum allowable service temperature of all components, which are less than or equal to

-40 °C (-40 °F).

Table 3.2.2-1 lists the maximum and/or minimum allowable temperatures for the critical cask components.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.2.2-1 Component Specifications - Minimum and Maximum Temperatures Material Min. Temp. Max. Temp. Reference

>1400°C 304/ 304L SS - Ref. 14 (Melting Temp.)

328°C Lead - Ref. 5 (Melting Temp.)

1093°C Polyurethane Foam - Ref. 15 (2000°F of Foam Char Temp.)

Seal (EPDM) -45°C 150°C Ref. 8 & 16 3.2.3 Content Properties As described in Chapter 1, Section 1.2.2.3 (Physical and Chemical Form - Density, Moisture Content and Moderators), the RT-100 is designed to transport contents that include contaminated resins/filters and activated hardware. The contents include secondary containers and may also include shoring. Resins are made of thermoplastics such as polystyrene, or material such as inorganic carbon or zeolite. Filters may be constructed from thermoplastics such as nylon, polyester, or polypropylene, or paper. Activated hardware are solid metallic components that are classified as high-density hardware or low-density hardware. Secondary containers are constructed of either coated/painted carbon steel or stainless steel, or a thermoplastic such as polyethylene or polypropylene. The filter media may be held within a stainless steel cartridge. Shoring can be made of wood or one or several of the materials comprising the secondary containers.

Based on the ASME code, Section II-D [Ref. 13], the acceptable temperature of the carbon steel and stainless steel material is approximately 525°C (977oF) for the range of loads and stresses occurring under NCT and HAC.

The melting temperatures of thermoplastics range from 100°C (212oF) up to 250°C (482oF) SFPE Handbook of Fire Protection Engineering, [Ref. 21], which typically soften at these temperatures and do not produce volatiles that could react with any of the contents. The auto-ignition temperature of thermoplastics is above 300°C (572oF) [Ref. 21].

The auto-ignition temperatures of paper and wood vary widely and are a function of their specific composition and moisture content. A commonly accepted value for the auto-ignition point for paper is 232°C (450°F) Fundamentals of Combustion Processes, [Ref. 22]. The auto-ignition point for wood has been shown to be at least 300°C (572oF) An Experimental Study of Autoignition of Wood, T. Poespowati, World Academy of Science, Engineering and Technology 23, 2008. [Ref. 23].

A summary of the maximum temperature specifications for the RT-100 contents is provided in Table 3.2.3-1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 3.2.3-1 Maximum Temperature Limits for RT-100 Content Materials Maximum Material Reference Temperature Carbon/Stainless Steel 525°C Ref. 13 Thermoplastics 300°C Ref. 21 Paper 232°C Ref. 22 Wood 300°C Ref. 23 3.3 Thermal Evaluation under Normal Conditions of Transport This section describes the thermal evaluations performed for the RT-100 for the NCT specified in 10 CFR 71.71 [Ref. 2]. The evaluation considers the response of the RT-100 to a range of temperature and environmental conditions as described in Section 3.3.1. The results are compared with allowable limits of temperature, pressure, etc., for the package components. The information is presented in summary tables, along with statements and appropriate comments. Information that is to be used in other sections of the review is identified. The margins of safety for package temperatures, pressures, and thermal stresses, including the effects of uncertainties in thermal properties, test conditions and diagnostics, and analytical methods are addressed.

The analyses are shown to be reliable and repeatable.

The following general information is considered and included in addressing the sections below, as appropriate:

o Assumptions that are used in the analysis are clearly described and justified.

o For computer analyses, including finite element analyses, the computer program is described and shown to be well benchmarked, widely used for thermal analyses, and applicable to the evaluation.

o Models and modeling details are clearly described.

o The methods used are properly referenced or developed in the application.

o These methods are correctly applied.

o The evaluation considers changes in package geometry and material properties resulting from structural and thermal tests under NCT and HAC.

o The required temperature and thermal boundary conditions for normal conditions of transport and hypothetical accident conditions are correctly applied.

o The time interval after the fire test is adequate to assure that maximum component temperatures and post-fire steady-state temperatures are achieved.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o The maximum temperatures and pressures of the components do not exceed their allowable values.

o Combustion of package components are considered, including the heat produced.

o Temperature data is reported at gaskets, valves, and other containment boundaries, particularly for temperature-sensitive materials as well as, for the overall package.

o Appropriate corrections and evaluations that account for differences in the thermal test are included for conditions like ambient temperature, decay heat of the contents, or package emissivity or absorptivity.

o Both interior and exterior temperatures are included.

o The damage caused by the tests and the results of any measurements made is reported in detail, including photographs of the testing and the test specimen.

3.3.1 Heat and Cold Section 3.3.1 demonstrates that the tests for NCT do not result in a significant reduction in the RT-100 effectiveness. The following items are considered and addressed:

o Degradation of the heat-transfer capability of the packaging (such as creation of new gaps between components) o Changes in material conditions or properties (e.g., expansion, contraction, gas generation, and thermal stresses) affecting structural performance o Changes in the packaging affecting containment, shielding, or criticality (such as thermal decomposition or melting of materials) o Ability of the packaging to withstand the tests under HAC The component temperatures and pressures are compared to their allowable values and do not exceed them. This section explicitly shows that the package meets the maximum temperature of the accessible package surface is less than 50 °C (122 °F) for non-exclusive-use shipment or 85 °C (185 °F) for exclusive use shipment when the package is subjected to the heat conditions of 10 CFR 71.43(g) [Ref. 2].

3.3.1.1 Load Cases Four load cases are analyzed in order to evaluate the RT-100 for the range of temperature and solar insolation conditions specified in 10 CFR 71.71 [Ref. 2] for normal conditions:

o Hot Case 1 o Hot Case 2 o Cold Case 1 o Cold Case 2 Robatel Technologies, LLC Page 3-14

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Hot case 1 is based on the requirements of 10 CFR 71.71(c)(1) [Ref. 2], which is one of the extreme initial conditions for normal conditions and a precursor for the hypothetical fire accident evaluation.

It has the following conditions:

o Ambient temperature, 38°C (100oF) o Initial temperature, 38°C (100oF) o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o Steady-state solar insolation, 776 W/m2 for flat surface and 388 W/m2 for curved surface o Internal heat load as a uniform heat flux, 13.04 W/m2 Hot case 2 is based on the requirements of 10 CFR 71.43(g) [Ref. 2] and has the following conditions:

o Ambient temperature, 38°C (100oF) o Initial temperature, 38°C (100oF) o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o No solar insolation, in shade o Internal heat load as a uniform heat flux, 13.04 W/m2 Cold case 1 is based on the requirements of 10 CFR 71.71(c)(2) [Ref. 2], which is another extreme initial condition for the NCT test evaluation. It has the following conditions:

o Ambient temperature, -40°C (-40oF) o Initial temperature, -40°C (-40oF) o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o No solar insolation, in shade o Internal heat load as a uniform heat flux, 13.04 W/m2 Cold case 2 is based on requirements of 10 CFR 71.71(b) [Ref. 2] and has the following conditions:

o Ambient temperature, -29°C (-20oF) o Initial temperature, -29°C (-20oF) o Heat transfer to ambient by natural convection, still air o Heat transfer to ambient by radiation o No solar insolation o Internal heat load as a uniform heat flux, 13.04 W/m2 Among them, Hot case 1 and Cold case 1 are two extreme conditions for the analyses. Hot case 1 is also referred to as the normal hot condition on which conservative boundary conditions are applied. This case provides the highest temperature distributions within the cask, and is used as initial conditions for evaluation of the hypothetical fire accident event as described in Section 3.4.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.3.1.2 Analytical Model The thermal evaluation of the RT-100 is performed using the ANSYS finite element computer software [Ref. 3]. The cask model is made of 3D thermal solid elements (SOLID90) that represent the major components of the cask. Contact between the lead and the inner and outer shells are modeled as bonded surfaces for thermal analyses in order to maximize heat input to the lead. The contact between the upper flange and the primary lid is modeled by a pair of 3D thermal contact elements (CONTA174) and 3D target elements (TARGE170), as are the other contacts between the primary lid and the secondary lid, the bolts with the primary lid, and the bolts with the secondary lid.

For conservatism, the top impact limiter is modeled as without the stainless steel plate covering the central hollow portion of the limiter. Thus, the concave area on the top impact limiter is exposed to solar insolation and/or fire. This approach leads to a conservatively high temperature over the top impact limiter. The contacts between the impact limiters and the cask body are also modeled by pairs of 3D thermal contact elements (CONTA174) and 3D target elements (TARGE170) between the relative surfaces.

A depiction of the ANSYS thermal model of the RT-100 is provided in Figure 3.3.1-1 and Figure 3.3.1-2. Additional details regarding the modeling and analysis of the RT-100 are presented in Calculation Package RTL-001-CALC-TH-0201, Rev. 6 [Ref. 4].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.3.1.3 Analysis Results The results of the steady state analyses of the cask model with impact limiters are presented in the form of temperature contour plots. Figure 3.3.1-3 through Figure 3.3.1-6 show the temperature contour plots for Hot case 1. Hot case 1 predicts the maximum temperatures experienced during NCT. The figures show the package, cask body, inner shell surface and lead shielding material, respectively. Figure 3.3.1-7 and Figure 3.3.1-8 provide the results for Hot case 2. Figure 3.3.1-9 and Figure 3.3.1-10 provide the results for Cold case 1. Cold case 1 represents the temperatures experienced by the package during extreme cold conditions. Figure 3.3.1-11 and Figure 3.3.1-12 provide the results for Cold case 2. Maximum temperature results are obtained by selecting the FE model component or material of interest and sorting the nodal results. Table 3.1.3-1 shows the maximum temperatures of the cask under NCT based on the steady state solution.

Figure 3.3.1-3 Temperature Contour Plot of PackageHot Case 1 Robatel Technologies, LLC Page 3-19

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-4 Temperature Contour Plot of Cask BodyHot Case 1 Robatel Technologies, LLC Page 3-20

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-5 Temperature Contour Plot of Inner Shell SurfaceHot Case 1 Robatel Technologies, LLC Page 3-21

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-6 Temperature Contour Plot of Lead ShieldingHot Case 1 Robatel Technologies, LLC Page 3-22

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-7 Temperature Contour Plot of PackageHot Case 2 Robatel Technologies, LLC Page 3-23

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-8 Temperature Contour Plot of Cask BodyHot Case 2 Robatel Technologies, LLC Page 3-24

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-9 Temperature Contour Plot of PackageCold Case 1 Robatel Technologies, LLC Page 3-25

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-10 Temperature Contour Plot of Cask BodyCold Case 1 Robatel Technologies, LLC Page 3-26

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-11 Temperature Contour Plot of PackageCold Case 2 Robatel Technologies, LLC Page 3-27

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 3.3.1-12 Temperature Contour Plot of Cask Body Cold Case 2 Robatel Technologies, LLC Page 3-28

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.3.2 Maximum Normal Operating Pressure The maximum pressure in the RT-100 for NCT is calculated using the maximum temperatures determined for the range of normal condition load cases. The calculation considers possible sources of gases including the following:

o Gases initially present in the package o Saturated vapor, including water vapor from the contents or packaging o Hydrogen or other gases resulting from thermal- or radiation-induced decomposition of materials such as water or plastics Summary of the pressure calculation is provided in the following sections. Additional details are provided in Calculation Package RTL-001-CALC-TH-0102, Rev. 6 [Ref. 6].

3.3.2.1 Calculation Method To determine the maximum normal operating pressure, the temperature of gas mixture within the cask is evaluated. Maximum temperature of the cask cavity under normal condition is bounded by the upper and lower temperature range of 80 °C (176oF) to -29 °C (-20oF). The total pressure in the cavity is represented by the sum of the primary contributors to the pressure. These are the pressure due to the increased temperature of the cavity gas (ideal gas law), the pressure due to the presence of water vapor, and the pressure due to the generation of gas via radiolysis.

The restriction of the contents to inorganic materials eliminates the potential for gas generation due to thermal degradation or biological activity. Thus, these gas sources are not considered in the evaluation. However, water vapor is present in trace quantities. Therefore, the analysis considers the contribution from the radiolytic decomposition of residual water in the cask cavity.

Per the ideal gas law, air pressure and water vapor pressure are directly proportional to the temperature, and with increase in temperature the pressure also increases. The upper bound temperature results in a higher maximum normal operating pressure for the cask compared to the lower bound. The gas mixture in the cavity is conservatively assumed to be at 80 °C (176oF).

3.3.2.2 Pressure Due to the Initially Sealed Air in the Cavity Per the ideal gas law, the partial pressure of the air (Pair) initially sealed in the fixed volume of the cask at the ambient temperature as it is heated to 80 °C (176 °F) is:

P1 x T2 = P2 x T1 Pair = 101.35 kPa[(353.15 K) / (294.25 K)] = 121.64 kPa (17.64 psia) 3.3.2.3 Pressure Due to the Water Vapor in the Cask The cask cavity is assumed to contain a small amount of water. By conservatively assuming a condensing surface temperature of 80 °C (176 °F), the water vapor pressure, Pwv, at this temperature is:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 47.34 kPa [6.87 psia], Fundamentals of Fluid Mechanics, Table B.2, pg. 831 [Ref. 17], Attachment 3.5-3.

Adding the water vapor pressure at 80 °C (176 °F) to the partial pressure of the air in the sealed cask at this temperature gives:

P2 = Pair + Pwv = 121.64 + 47. 34 = 169.0 kPa

= 169.0 kPax 0.145 psi/kPa = [24.51 psia]

3.3.2.4 Pressure Due to Generation of Gas Solidified or dewatered material may contain some water. Therefore, radiolytic generation of gases from this water could occur. Hydrogen and oxygen may be produced in the cask by radiolytic decomposition of residual water in the cask contents. As described in Chapter 1, Section 1.2.2.6, the maximum quantity of hydrogen must be limited to less than 5% to ensure that an explosive quantity does not accumulate.

The cask atmosphere can be assumed to contain 5% of hydrogen (H2) gas due to radiolysis of the water. By stoichiometry of the water molecule (H2O), the cask atmosphere will also contain 2.5%

oxygen (O2) gas generated by radiolysis. Partial pressures in an ideal gas mixture are additive and behave the same as ideal gas volume fraction or mole fractions. Therefore, the partial pressure of hydrogen is described by the following equation:

PH2 = 0.05 Ppt Where, Ppt = Pair + Pwv + PH2 + PO2 Combining Pair + Pwv = P2 and noting that PO2 = 0.5 x PH2.

PH2 = 0.05 x (P2 + 1.5 PH2)

Solving the equation explicitly for PH2 give:

PH2 = [0.05 P2] / [1 - 0.05 (1.5)]

= [0.05

169.0 kPa] / [1 - 0.05 (1.5)]

= 9.14 kPa [1.32 psia]

3.3.2.5 Total Pressure Based on the stoichiometric relationship between hydrogen and oxygen liberated by radiolysis of water, and again combining the pressure of the initially sealed air and water vapor as P 2, the total pressure in the cask at 80oC (176oF) is:

PTotal = P2 + 1.5 PH2

= 169.0 kPa + 1.5

9.14 kPa

= 182.71 kPa [26.5 psia] (actual calculated MNOP)

The design basis maximum normal operating pressure (MNOP) value is conservatively set at 342.7 kPa (49.7 psia) for use in the cask structural analyses for NCT.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.4 Thermal Evaluation under Hypothetical Accident Conditions This section describes the thermal evaluation of the RT-100 under hypothetical accident conditions.

The RT-100 is evaluated by finite element computer analysis rather than physical testing to demonstrate the performance of the cask in response to the fire test conditions specified in 10 CFR 71.73(c) [Ref. 2]. The HAC defined in 10 CFR 71.73(c) [Ref. 2] are applied sequentially, considering the damaged condition of the packaging following the 30-foot free drop and pin puncture accident events prior to the fire transient. For the accident condition thermal evaluation, the general comments in Section 3.3 are considered and addressed, as appropriate.

As described in Chapter 2, Section 2.7.3 (Puncture), different pin puncture configurations are considered in order to determine the worst case for the accident event. For the structural evaluation, the orientations considered are directly in the middle of the secondary lid to maximize the bending loads in the primary and secondary lids and prying forces in the bolts. The second configuration considers the pin impact directly into the side of the RT-100 to ensure that the outer shell is not punctured by the pin. For the thermal analysis, these two events are also considered to be limiting.

The differences are in the location of the pin at impact.

3.4.1 HAC Fire AnalysisPin Puncture Damage to Top Impact Limiter The analytical model described in Section 3.3.1.2 is used to evaluate the RT-100 package with damage on the top impact limiter. For this case, the limiting configuration for the thermal analysis considers a pin puncture through the top impact limiter directly into the secondary lid at the location of the O-rings. The model placed a 150 mm (6 in) diameter hole through the upper impact limiter, directly exposing the secondary lid to the thermal environment of the hypothetical accident fire. The following section evaluates both pin puncture orientations to determine the effect to critical components such as the seal locations and lead shielding.

3.4.1.1 Initial ConditionsPin Puncture Damage to Top Impact Limiter Per Regulatory Position 1.1 in Regulatory Guide 7.8 [Ref. 20], the initial cask temperature distribution is considered to be at steady state with an ambient temperature of 38°C (100oF) and solar insolation prior to the HAC fire accident. To meet this requirement, the steady-state solution for NCT hot case 1 is used, obtained to the initial temperatures of the cask prior to the fire. The steady-state temperatures are applied as the first load step of the transient solutions. To account for damage to the package that results during the sequential drop accidents, damage due to pin puncture is considered during the top and side puncture. For the top impact limiter 150 mm diameter volume of material including the steel shell and FR3740 foam is removed at the point closest to the elastomer O-ring. This is a conservative approach since the puncture probe will not penetrate through the top skin of the impact limiter and compressed foam will remain beneath the point of impact. Figure 3.4.1-1 shows the temperature contour of the package prior to the fire accident and localized higher temperatures in the region of the damaged impact limiter. Figure 3.4.1-2 and Figure 3.4.1-3 show the cask body and inner shell cavity temperature distribution prior to the fire accident.

3.4.1.2 HAC Fire and Cool-down AnalysisPin Puncture Damage to Top Impact Limiter The thermal analysis for HAC includes a 30 minute transient fire followed by the prescribed post-fire cool-down period. The FE model described in Section 3.3.1.2 is analyzed by applying the Robatel Technologies, LLC Page 3-31

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 following boundary conditions. Following the initial load step in which the steady-state temperatures are applied, the analysis proceeds with the HAC fire transient for 30 minutes (1,800 seconds) followed by a cool-down period with the boundary conditions associated with NCT Hot case 1. The NCT Hot case 1 boundary conditions are applied as constants ignoring the day/night cool-down cycle. The following is a summary of the fire transient boundary conditions:

o Environment temperature, 800°C (1472°F) o No solar insolation, 0 W/m2 o Forced convection, heat transfer coefficient = 10 W/m2°C o Radiation from the environment to package surface, flame emissivity = 0.9 o Internal heat load as a uniform heat flux, 13.04 W/m2 The cool-down analysis is performed for 216,000 seconds (2.5 days) with the following boundary conditions:

o Environment temperature, 38°C (100°F) o Solar insolation applied as constant, 776 W/m2 for flat surfaces and 388 W/m2 for curved surfaces.

o Natural convection, heat transfer coefficient = 5 W/m2°C o Radiation from package surface to the environment, package emissivity = 0.8 o Internal heat load as a uniform heat flux, 13.04 W/m2 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b) 3.4.2 HAC Fire EvaluationPin Puncture Damage to the Side of the Cask Body The analytical model described in Section 3.3.1.2 is used to evaluate the RT-100 with damage on the cask side wall. For this case, the limiting configuration considers the pin puncturing the thermal shield directly below the lifting block. This configuration increases the area of the outer shell of the cask that is not protected by the thermal shield and maximizes the heat input into the lead. The following section evaluates both pin puncture orientations to determine the effect to critical components such as the seal locations and lead shielding.

3.4.2.1 Initial ConditionPin Puncture Damage to the Side of the Cask Body Figure 3.4.2-1 shows the FE model of the cask body due to pin puncture damage on the side. The location of the damage is chosen below the lifting pocket in a region where no thermal insulation exists. Therefore, the heat flow is maximized into the package. The removed elements at this area cover a surface area greater than the area of the pin. As with the pin puncture on the top impact limiter case, NCT Hot case 1 steady state solution is used as the initial condition for the fire cases.

The steady-state temperatures are applied as a boundary condition during the first load step of the Robatel Technologies, LLC Page 3-43

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 transient solution, prior to initiating the HAC fire transient. Figure 3.4.2-2 shows the package temperature distribution prior to the fire. The cask body and inner shell pre-fire temperatures are shown in Figure 3.4.2-3 and Figure 3.4.2-4.

3.4.2.2 HAC Fire AnalysisPin Puncture Damage to the Side of the Cask Body The thermal analysis for HAC includes a 30 minute transient fire followed by the prescribed post-fire cool-down period. The FE model described in Section 3.3.1.2 is analyzed by applying the following boundary conditions. Following the initial load step in which the steady-state temperatures are applied, the analysis proceeds with the HAC fire transient for 30 minutes (1,800 seconds) followed by a cool-down period with the boundary conditions associated with NCT hot case 1. The NCT hot case 1 boundary conditions are applied as constants, ignoring the day/night cool-down cycle. The following is a summary of the fire transient boundary conditions:

o Environment temperature, 800°C (1472°F) o No solar insolation, 0 W/m2 o Forced convection, heat transfer coefficient = 10 W/m2°C o Radiation from the environment to package surface, flame emissivity = 0.9 o Internal heat load as a uniform heat flux, 13.04 W/m2 The cool-down analysis is performed for 216,000 seconds (2.5 days) with the following boundary conditions:

o Environment temperature, 38°C (100°F) o Solar insolation applied as constant, 776 W/m2 for flat surfaces and 388 W/m2 for curved surfaces.

o Natural convection, heat transfer coefficient = 5 W/m2°C o Radiation from package surface to the environment, package emissivity = 0.8 o Internal heat load as a uniform heat flux, 13.04 W/m2 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.4.3 Maximum Temperatures and Pressure This section summarizes the peak accident condition temperatures of RT-100 components as a function of time both during and after the fire, as well as the maximum temperatures from the post-fire, steady-state condition. This section includes those temperatures at locations in the package that are significant to the safety analysis and review. The calculations of transient temperatures trace the temperature-time history up to and past the time at which maximum temperatures are achieved and begin to fall. The calculations confirm that these temperatures do not exceed their maximum allowable values. It also confirms that lead shielding does not reach melting temperature.

The RT-100 is evaluated structurally for the maximum HAC temperatures and pressures in Chapter 2, Section 2.7.4 (Thermal).

3.4.3.1 Maximum Temperatures Section 3.4.1 and 3.4.2 present a summary of the evaluation of the RT-100 for the hypothetical accident condition fire transient. Provided in the summary are figures depicting temperature distributions and time histories as a function of time during and after the fire transient. Maximum temperatures for various cask components as a result of the HAC are presented in Table 3.1.3-2 and Table 3.1.3-3.

Of interest in this section is the determination of the maximum internal pressure in the cask cavity as a result of the fire test. As shown in Table 3.1.3-3, the maximum average inner shell temperature during the fire transient is 137°C. The temperature of the cask body components is increased due to the fire transient, with a maximum normal condition inner shell temperature of 73.1°C as reported in Table 3.1.3-1. Because the temperature of the inner shell of the cask is raised by 64°C as a result of the fire transient and because the maximum internal decay heat of the contents is only 200 watts, it is conservative to assume that the cavity temperatures are bounded by the average inner shell temperatures. For conservatism, the inner shell can be assumed to be at 150°C for the pressure calculations presented in Section 3.4.3.2.

As previously discussed, the primary components of interest during the fire transient from a temperature standpoint are the lead gamma shielding and the O-ring seals in the primary and secondary lids. As described in detail in Section 3.4.2, the lead and O-ring materials do not exceed their allowable, and in fact have safety margins of more than 23°C below their maximum allowable temperatures. The temperature distributions within the cask, as a result of the hypothetical accident condition fire transient, are fully considered in the structural evaluation of the cask presented in Chapter 2, Section 2.7.4 (Thermal).

3.4.3.2 Maximum Accident Condition Pressure The evaluation of the maximum pressure in the RT-100 is based on the maximum normal operating pressure, and considers fire-induced increases in package temperatures, thermal combustion or decomposition processes, phase changes, etc. (Fuel rod failure is not applicable). The value of this maximum pressure is consistent with the values used in the Structural Evaluation and Containment sections.

Similar to the calculation of the maximum normal operating pressure in Section 3.3.2, the Robatel Technologies, LLC Page 3-57

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 maximum accident condition pressure is calculated using bounding assumptions for the temperatures in the cask as a result of the hypothetical accident condition fire transient. The maximum pressure is the sum of four components:

1. The pressure due to the initially sealed air in the cavity

2. The pressure due to water vapor in the cask

3. The pressure due to the hydrogen and oxygen gases generated by radiolysis

4. The pressure due to the thermal decomposition of the contents The following sections present a summary of the maximum accident condition pressure calculation.

Details of the calculation are provided in Calculation Package RTL-001-CALC- TH-0202, Rev. 6

[Ref. 7], and RTL-001-CALC-TH-0301, Rev. 1 [Ref. 25].

3.4.3.2.1 Calculation Method The internal cavity pressure due to accident condition temperatures is determined using the same method used to calculate the maximum normal condition pressure in Section 3.3.2. The method presented below is equal to that used previously, with the maximum normal operating pressure and internal temperatures used along with the maximum internal temperature determined in Section 3.4.3.1 to calculate the maximum accident condition pressure.

3.4.3.2.2 Pressure Due to the Initially Sealed Air in the Cavity Per the ideal gas law, the partial pressure of the air (Pair) initially sealed in the fixed volume of the cask at the ambient temperature as it is heated to 150 °C is:

P1 x T2 = P2 x T1 Pair = 101.35 kPa[(423.15 K) / (294.25 K)] = 145.8 kPa (21.15 psia) 3.4.3.2.3 Pressure Due to the Water Vapor in the Cask The RT-100 cavity is assumed to contain a small amount of water. By conservatively assuming a condensing surface temperature of 150 °C, the water vapor pressure, Pwv, at this temperature is 475.8 kPa [69 psia] Fundamentals of Engineering Thermodynamics, 5th Edition, Table A-2 on pg.

761 [Ref. 18], also see Attachment 3.5-4. Adding the water vapor pressure at 150 °C to the partial pressure of the air in the sealed cask at this temperature gives:

P2 = Pair + Pwv = 145.8 + 475.8 = 621.6 kPa [90.16 psia]

3.4.3.2.4 Pressure Due to Generation of Gas Solidified or dewatered material may contain some water. Therefore, radiolytic generation of gases from this water could occur. Hydrogen and oxygen may be produced in the cask by radiolytic decomposition of residual water in the cask contents. As described in Section 1.2.2.6, the maximum quantity of hydrogen must be limited to less than 5% to ensure that an explosive quantity does not accumulate.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The cask atmosphere can be assumed to contain 5% of hydrogen (H2) gas due to radiolysis of the water. By stoichiometry of the water molecule (H2O), the cask atmosphere will also contain 2.5%

oxygen (O2) gas generated by radiolysis. Partial pressures in an ideal gas mixture are additive and behave the same as ideal gas volume fraction or mole fractions. Therefore, the partial pressure of hydrogen is described by the following equation:

PH2 = 0.05 Ppt Where, Ppt = Pair + Pwv + PH2 + PO2 Combining Pair + Pwv = P2 and noting that PO2 = 0.5 x PH2.

PH2 = 0.05 x (P2 + 1.5 PH2)

Solving the equation explicitly for PH2 gives:

PH2 = [0.05 P2] / [1 - 0.05 (1.5)]

= [0.05

621.6kPa] / [1 - 0.05 (1.5)]

= 33.6 kPa [4.87 psia]

3.4.3.2.5 Total Pressure Based on the stoichiometric relationship between hydrogen and oxygen liberated by radiolysis of water, and again combining the pressure of the initially sealed air and water vapor as P2, the total pressure in the cask at 150 °C is:

PTotal = P2 + 1.5 PH2

= 621.6 kPa + 1.5

33.6 kPa

= 672 kPa [97.47 psia]

The maximum pressure is 672 kPa [97.47 psia] under HAC. For conservatism, the maximum accident pressure is assumed to be 689.4 kPa [100 psia] for the structural analyses presented in Chapter 2, Section 2.7.4 (Thermal).

3.4.3.2.6 Total Pressure Accounting for Combustion of Contents In addition to the natural effect of temperature increases on pressure buildup in the package, other thermally driven phenomena can contribute to the pressure buildup within the containment boundary of a package. As discussed previously, these include phase transformation of materials in the package and radiolysis of the contents by radioactive decay. Additionally, the pressure increases due to the contribution of the partial pressure that results from the thermal decomposition of the package contents [Ref 25].

Solid polymeric materials, including cellulosics such as wood and paper, undergo both physical and chemical changes when heat is applied. Thermal decomposition is a process of extensive chemical species change caused by heat, generating gaseous fuel vapors which can burn above the Robatel Technologies, LLC Page 3-59

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 solid material. The process is self-sustaining when the burning gases feed back sufficient heat to the material to continue the production of gaseous fuel vapors or volatiles. These volatiles react with the oxygen in the air to generate heat, and part of this heat is transferred back to the polymer to continue the process.

The Robatel RT-100 contents include filters that may be constructed from thermoplastics (nylon, polyester, polypropylene) or paper and shoring made of wood may be contained in the package.

Although it is unlikely that temperatures under HAC will approach the auto-ignition temperatures of the contents, the following analysis is performed to evaluate the effect of combustion on the package pressure.

Combustion in a sealed container is limited by the amount of air present to support the chemical reaction for the thermal decomposition of the fuel. Heats from the exothermic combustion reaction will increase the temperature of the contents and packaging. The maximum temperature in a sealed container will determine the maximum pressure, along with some additional pressure from emitted gases. The sealed inner containment of the RT-100 cask contains only enough air (5.75 kg) for complete combustion of approximately 1.127 kg of cellulosic material, paper or wood; or 0.390 kg of polyethylene.

Gibbs-Dalton Law defines total pressure, PT, equal to the sum of the partial pressures of the individual gases present. The total pressure PT, in the package containment is the sum of pressures due to phase transformation of materials in the package Pv (Ref. 25, p. 25, where Pv = Psat),

radiolysis of the contents by radioactive decay Pr (1.5 PH2 from Section 3.4.3.2.5), and thermal decomposition of the package contents Pf (Ref. 25, p. 25, where Pf = Pfwood). The vapor pressure from the phase transformation of water and the partial pressures of hydrogen and oxygen gases generated from the radiolysis of water in the contents are considered in the total pressure calculation.

PT = Pv + Pr + Pf PT = 463.2 kPa + 50.4 kPa + 171.0 kPa = 684.6 kPa [99.3 psia]

where the total pressure of the inner cavity is based on the complete combustion of wood, which has the highest heat of combustion. Since the temperature required to ignite wood are not sustainable, complete combustion is not considered a credible event, therefore, the maximum pressure is taken as 97.47 psia as demonstrated in Section 3.4.3.2.5.

3.4.4 Maximum Thermal Stress The RT-100 cask is evaluated for the stresses produced by the temperature gradients in the cask body that result from exposure of the cask to the HAC fire transient. This evaluation, which utilizes the temperature distributions resulting from the fire accident as described in Section 3.4.3, is presented in detail in Chapter 2, Section 2.7.4 (Thermal).

3.4.5 Accident Conditions for Fissile Material Packages for Air Transport This Section is NOT APPLICABLE. The RT-100 is not be used for fissile material air transport.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.5 Appendix Attachment 3.5-1 EPDM Temperature Specifications

[Ref. 16]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 3.5-2 Seal Material EPDM Working Temperature

[Ref. 8]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 3.5-3 Water Vapor Pressure Reference (80°C)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 3.5-4 Water Vapor Pressure Reference (150°C)

[Ref. 18]

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[Ref. 18]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 3.5-4 Water Vapor Pressure Reference (150°C) (Continued)

[Ref. 18]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 3.6 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL, dated March 7, 2012

3. ANSYS, Release 14.0, ANSYS, Inc., Canonsburg, PA, October, 2011

4. RTL-001-CALC-TH-0201, Rev. 6, "RT-100 Cask Thermal Analyses" (PROPRIETARY)

5. Fundamentals of Heat and Mass Transfer, Frank P. Incropera, David P. DeWitt, 2002, 5th ed., John Wiley & Sons, Inc.

6. RTL-001-CALC-TH-0102, Rev. 6, "RT-100 Cask Maximum Normal Operating Pressure Calculation" (PROPRIETARY)

7. RTL-001-CALC-TH-0202, Rev. 6, "RT-100 Cask Hypothetical Accident Condition Maximum Pressure Calculation" (PROPRIETARY)

8. TRELLEBORG Sealing Solutions O-Ring and Backup Rings Catalog, August 2011 Edition

9. UNIFRAX Fiberfrax 970 Ceramic Paper Data Sheet Proprietary Information Content Withheld Under 10 CFR 2.390(b)

13. ASME Boiler & Pressure Vessel Code 2007 Edition,Section II - Part D, "Materials",

The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

14. Sanghavi Bothra Engineering Co. Pvt. Ltd.(SBE) 304/304L Stainless Steel Product Mechanical and Physical Properties

15. GENERAL PLASTICS Design Guide for LAST-A-FOAM FR-3700 Crash & Fire Protection of Radioactive Material Shipping Containers, Rev. 02.20.12

16. Parker O-Ring Handbook ORD 5700, Retrieved on August 28, 2013, Retrieved from http://www.parker.com/literature/ORD%205700%20Parker_O-Ring_Handbook.pdf.

17. Fundamentals of Fluid Mechanics, B. Munson, D. Young and T. Okiishi, 4th ed., John Wiley

& Sons, Inc.

18. Fundamentals of Engineering Thermodynamics, M. Moran and H. Shapiro, 5th ed., John Wiley & Sons, Inc.

19. Glenn Lee, Radiation Resistance of Elastomers, IEEE Transactions on Nuclear Science, Vol.

NS-32, No. 5, October 1985

20. U.S. Nuclear Regulatory Commission, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, Regulatory Guide 7.8.

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21. SFPE Handbook of Fire Protection Engineering, "Thermal Decomposition of Polymers," C.L. Hirschler, M. Marvelo, Chapter 7 of 3rd Edition, NFPA, 1 Batterymarch Park, Quincy, MA, 2001, www.nfpa.org.

22. Fundamentals of Combustion Processes, A. McAllister, J. Chen, A. Fernandez-Pello, Springer, 2011.

23. An Experimental Study of Autoignition of Wood, T. Poespowati, World Academy of Science, Engineering and Technology, Vol. 23, 2008., Retrieved on August 28, 2013, Retrieved from http://www.waset.org/journals/waset/v23/v23-13.pdf.

24. ASME Boiler & Pressure Vessel Code, 2010,Section II, Part D, Materials, The American Society of Mechanical Engineers, New York, NY 2010

25. RTL-001-CALC-TH-0301, Rev. 1, RT-100 Cask Hypothetical Accident Condition Combustion Analysis (PROPRIETARY)

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4. CONTAINMENT EVALUATION Robatel performed the containment evaluation of the RT-100 using the Nuclear Industry standards and under the RT Company Quality Assurance Program [Ref. 1]. This Chapter demonstrates the RT-100 containment boundary compliance with the permitted activity release limits specified in 10 CFR 71.51(a)(1) [Ref. 2] and 10 CFR 71.51(a)(2) [Ref. 2] for both normal conditions of transport (NCT) and hypothetical accident conditions (HAC) of transport. The reference leakage rates for various cask conditions are normally calculated, and the most bounding value is chosen as the maximum allowable leakage rate for the cask in order to ensure compliance with regulatory limits.

Due to the variety of inventories, diversity in both isotopic composition and in total activity concentration, the RT-100 has been established as a leaktight container. Leaktight is a degree of package containment that in a practical sense precludes any significant release of radioactive materials.

This degree of containment is achieved by demonstration of a leakage rate less than or equal to 1 x 10-7 ref*cm3/s, of air at an upstream pressure of 1 atmosphere absolute and a downstream pressure of 0.01 atmosphere absolute or less (ANSI N14.5-2014 [Ref. 3]).

The containment review is based in part on the descriptions and evaluations presented in the General Information, Structural Evaluation and Thermal Evaluation sections of the application. Similarly, results of the containment review are considered in the review of Operating Procedures and Acceptance Tests and Maintenance Program. An example of the information flow for the containment review is shown in Figure 4-1 on the following page.

4.1 Description of Containment System Section 4.1 provides a detailed description of the containment system. This description includes the containment vessel, welds, seals, lids, cover plates, and other closure devices relevant to the containment boundary of the cask. Materials of construction and applicable codes and standards are presented in the RT100 NM 1000 Rev. G - Bill of Material (Chapter 1, Appendix 1.4, Attachment 1.4-1).

4.1.1 Containment Vessel The package containment system is defined as the inner shell of the shielded transport cask, together with the associated lid, O-ring seals, and lid closure bolts. The inner shell of the RT-100, or containment vessel, consists of a right circular cylinder of 1730 mm inner diameter and 1956 mm inside height. The shell is fabricated of stainless steel. At the base, the cylindrical shell is attached to a circular forged bottom with full penetration weld. At the top, the inner shell is attached to a circular forged flange with a full penetration weld. The primary lid is attached to the cask body with thirty-two (32) equally spaced M48 hex head bolts. A secondary lid covers an opening in the primary lid and is attached to the primary lid using eighteen (18) equally spaced M36 hex head bolts. Refer to Chapter 4, Section 4.1.4 for closure details. The inner shell is shown to maintain stresses within allowable limits in Chapter 2, Section 2.6.7 for NCT and in Chapter 2, Section 2.7 Robatel Technologies, LLC Page 4-1

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 for HAC. These evaluations demonstrate that the inner shell maintains its integrity and provides containment along with the closure system as described in Section 4.1.4.

Figure 4-1 Information Flow for the Containment Review Structural General Thermal Evaluation Information Evaluation

Deformation

Dimensions

Temperatures

Chemical and

Contents

Pressures Galvanic Reactions

Materials

Gas Inventory

Contents Condition

Containment Boundary Containment Review Containment Boundary Normal Conditions of Hypothetical Accident Transport Conditions

Components

Criterion *Gamma

Penetrations

Demonstration of *Neutron

Closure Compliance

Seals

Combustible Gases Operating Acceptance Tests Procedures and Maintenance

Closure

Fabrication Requirements Verification

Assembly Leakage Rate Verification

Periodic Verification Leakage Rate Robatel Technologies, LLC Page 4-2

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.1.2 Containment Penetration There are three locations where the containment vessel may be penetrated. For each location, an inner O-ring seals the containment boundary.

o Primary lid o Secondary lid o Cask vent port cover plate A vent port penetrates the primary lid into the main cask cavity. The vent penetration contains a quick disconnect valve and is sealed with the vent port cover plate. The primary lid, secondary lid and the cover plate are sealed with EPDM O-rings. Figure 4.1.2-1 illustrates the containment boundaries of the RT-100 (in red). The RT-100 does not rely on any valve or pressure relief device to meet the containment requirements. The quick disconnect valve is protected by the vent port cover plate which protects the valve from unauthorized operation and provides a sealed enclosure to retain any leakage from the device.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.1.3 Welds and Seals The containment vessel is fabricated using full penetration welds. Lid seals are EPDM O-rings and are further addressed in Section 4.1.4. O-rings may be supplied by manufacturers such as those in Parker O-Ring Handbook ORD 5700 [Ref. 10] and Trelleborg Sealing Solutions O-Ring and Backup Rings Catalog, August 2011 [Ref. 11]. Additional information on the O-rings taken from these references is provided in Attachment 4.5-1 through Attachment 4.5-5.

Parker O-Ring Handbook ORD 5700 [Ref. 10] and Trelleborg Sealing Solutions O-Ring and Backup Rings Catalog, August 2011 [Ref. 11] contain information regarding the operating temperature range, gap permeability, and compression set for the material. The temperature performance of the EPDM O-rings is presented in Chapter 3, Section 3.2.1 and the application of the O-rings in the primary and secondary lid seals is addressed in Chapter 2, Appendix 2.13. EPDM radiation resistance is addressed in Radiation Resistance of Elastomers, IEEE Transactions on Nuclear Science, Vol. NS-32, No.5, October 1985 [Ref. 12], indicating that the material is radiation resistant up to 5x108 rads while retaining reasonable flexibility and strength, hardness, and very good compression set resistance. A copy of Reference 12 is provided in Attachment 4.5-5.

4.1.4 Closure The primary lid closure consists of a partially recessed, 210 mm-thick stainless steel plate. The lid is supported at the perimeter of the cylindrical body by a thick flange (upper forging) which is welded to the top of the inner and outer cylindrical shells. The Primary Lid is attached to the cask body by thirty-two (32) equally spaced M48 hex head bolts. Two (2) EPDM O-rings are retained in machined grooves at the lid perimeter. Groove dimensions prevent over-compression of the O-rings by the closure bolt pre-load forces and hypothetical accident impact forces.

The cask is fitted with a recessed secondary lid which consists of 100 mm thick plate, a 60 mm thick lead gamma shield, and 10 mm thick closure plate. The Secondary Lid is attached to the Primary Lid with eighteen (18) equally spaced M36 hex head bolts. Two (2) EPDM O-rings are retained in machined grooves at the lid perimeter.

The quick-disconnect valve is housed under a 10mm thick stainless steel cover plate. The Quick-Disconnect Valve Cover Plate is attached to the primary lid with six (6) equally spaced M10 hex head bolts. Two (2) EPDM O-rings are retained in machined grooves at the lid perimeter.

The torque requirements for these bolts may be seen below in Table 4.1.4-1. Due to this closure setup, continuous venting from the RT-100 is precluded.

As stated above, the containment system is sealed by multiple bolted closures. These closures contain numerous bolts that are required to be tightened to specified torques using approved procedures during the cask loading process. Secure closure is assured by the torque values specified Robatel Technologies, LLC Page 4-5

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 and the assembly verification leak test performed prior to transport. The torques are specified in order to ensure that sufficient pre-load is applied to the bolts so that they will withstand loads from the maximum normal and accident condition pressures within the cavity.

The closure system is evaluated for NCT and HAC in Chapter 2, Appendix 2.13. Closure bolts are shown to maintain adequate design margin and allow the O-rings to maintain a positive seal at all times.

Table 4.1.4-1 Bolt Torque Requirements Torque Values (N-m)

Location Size

+/- 10% Lubricated Primary Lid M48 850 Secondary Lid M36 350 Quick-Disconnect Valve M10 27 Cover Plate 4.1.5 Cavity Volume, Conditions, and Contents The cavity dimensions are displayed in Table 4.1.5-1.

Table 4.1.5-1 Cask Cavity Dimensions Inches Centimeters Lcavity 77.2 196 Dcavity 68.1 173 Thus, the volume of the cylindrical cavity is Vcavity = (*Dcavity2*Lcavity)/4 Table 4.1.5-2 Cask Cavity Volume Total Cavity Volume [cm3] 4.60E+06 The temperatures under normal and accident conditions are determined based on the maximum internal cavity temperatures for normal and accident situations. Pressures and temperatures are provided by Calculation Package RTL-001-CALC-TH-0102, Rev. 6 [Ref. 8] and Calculation Package RTL-001-CALC-TH-0202, Rev. 6 [Ref. 9] for normal and accident situations, respectively. The standard leakage rate is the leakage rate of dry air when it is leaking from 1 atm (upstream pressure) to 0.01 atm (downstream pressure) at 298 K (ANSI N14.5-2014 [Ref. 3]).

Dynamic viscosity values were generated based on the Sutherland equation (Fundamentals of Fluid Mechanics, 5th edition [Ref. 14]), Introduction to Nuclear Engineering, 3rd edition [Ref.13]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table IV.4, ANSI N14.5-2014 [Ref. 3] Table B-1, Fundamentals of Fluid Mechanics Table B.4

[Ref. 14] Table B.4, and viscosity of gaseous helium table in Brookhaven National Laboratory, Selected Cryogenic Data Notebook, Aug. 1980 [Ref. 15].

Table 4.1.5-3 Parameters for Normal Transport and Accident Conditions Parameter Normal Conditions Accident Conditions Standard Conditions Pu [atm] 3.38 6.8 1 Pd [atm] 1 1 0.01 Pa [atm] 2.19 3.9 0.505 T [°F] 176 (353 K, 80 °C) 302 (423 K, 150 °C) 76.7 (298 K, 25 °C)

M [g/mol] 29 (air), 4 (He) 29 (air), 4 (He) 29 (air), 4 (He)

[cP] 0.0207 (air), 0.0224 (He) 0.0236 (air), 0.0254 (He) 0.0185 (air), 0.0198 (He) a [cm] 0.49 0.49 0.49 4.2 Allowable Leakage Rates at Test Conditions Un-choked flow correlations are used as they better approximate the true measured flow rate for the leakage rates associated with transportation packages. Using the equations for molecular and continuum flow provided in NUREG/CR-6487 [Ref. 7], the corresponding leak hole diameter is calculated for the RT-100 for standard test conditions by solving Equation 4.1 for D, the leak hole diameter. The capillary length required for Equation 4.1 for the containment system is conservatively chosen as the O-ring groove width in the vent port cover plate lid, which is 0.49 cm.

Equation 4.1 3 3 T 2.49x106 D4 3.81x10 D M L@a = + x [Pu Pd ]

a aPa

[ ]

where:

L@Pa is the allowable leakage rate at the average pressure for standard conditions [cm3/s],

a is the capillary length [0.49 cm],

T is the temperature for standard conditions [K],

M is the gas molecular weight [g/mol] = 29.0 for air, 4.0 for He from ANSI N14.5-2014, Table B1, is the dynamic viscosity for helium or air [cP],

Pu is the upstream pressure [atm],

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Pd is the downstream pressure [atm],

Pa is the average pressure; Pa = (Pu + Pd)/2 for standard conditions [atm], and D is the capillary diameter [cm].

The leak hole diameter is determined using the parameters for standard conditions presented in Table 4.1.5-3.

The allowable leakage rate for leaktight conditions is at the upstream pressure, the ratio presented in Equation 4.2 is used to convert Equation 4.1 to upstream leakage rate so that the capillary diameter can be determined.

Equation 4.2 Pu

@ = L@Pu Pa where:

L@Pa is the allowable leakage rate at the average pressure [cm3/s] for standard conditions, L@Pu is the allowable leakage rate at the upstream pressure [cm3/s] for standard conditions, Pu is the upstream pressure [atm],

Pd is the downstream pressure [atm], and Pa is the average pressure; Pa = (Pu + Pd)/2 [atm].

The sensitivity for the leakage test procedures is established by ANSI N14.5-2014 [Ref. 3] as shown in Equation 4.3.

Equation 4.3 S = 1/2*Leakage Rate1 4.3 Leakage Rate Test for Type B Packages This section describes the leakage tests used to show that the RT-100 meets the containment requirements of 10 CFR 71.51 [Ref. 2]. Leak test requirements are further specified in Chapter 8, Section 8.1.4.

The following leakage tests are conducted on the RT-100 as required by ANSI N14.5-2014 [Ref.

3]:

1 Leakage rate in this case is the upstream pressure leakage rate at standard conditions.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.1.5-1 Leakage Tests of the RT-100 Package Test Frequency Test Gas Acceptance Criteria Maintenance After maintenance, repair (such as weld repair), or replacement of components of the containment Helium LHe system Fabrication Prior to the first use of the RT-100 Periodic Within 12 months prior to next shipment Pre-Shipment Before each shipment of Type B Nitrogen *No Leakage at waste or air a Sensitivity (optional) 10-3 ref-cm3/sec

Adjusted for the individual properties of the test gas (calculated below); sensitivity is LHe/2.

As shown in Table 4.1.5-1, the Maintenance, Fabrication, and Periodic leakage tests may be performed using helium as the test gas. The acceptance criterion for these tests is the equivalent reference leakage rate for helium gas, LHe, which is calculated below.

4.3.1 Determination of Equivalent Reference Leakage Rate for Helium Gas Section 4.3.1 determines the allowable leakage rate using the Helium gas which may be used to perform the annual verification leakage tests summarized in Table 4.1.5-1 above. This calculation uses formulas presented in ANSI N14.5-2014 [Ref. 3].

It is known that the reference air leakage rate, LR, is 1.00 x 10-7 ref*cm3/s based on leaktight criteria.

Using Equation 4.1 and Equation 4.2, the maximum capillary diameter, Dmax, was determined:

298 3.81103 3 29 2.49106 4

0.505

@= + (1 0.01) ( )

(0.49)(0.0185) (0.49)(0.505) 1

( )

= 1107 3 /

Diameter values are inputted until the result of the above calculation is roughly equivalent to 1x10-7 ref*cm3/s. Solving for Dmax iteratively yields:

Dmax = 1.3261E-04 [cm]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The equivalent air/helium mixture that would leak from Dmax during a leak test, as described in Table 4.1.5-3, is determined. The leakage tests are performed with an air/helium mixture. The helium partial pressures can vary from 0.25 atm to 1.0 atm. An example with a helium partial pressure of 0.7 atm has been provided to illustrate the process used to determine the value of the variables used to determine the acceptable test leakage rates.

Assume the cask void is evacuated to 0.3 atm and then pressurized to 1.0 atm with an air/helium mixture.

Pvoid = Pair = 0.3 atm Pmix= 1.0 atm PHe = Pmix - Pair = 0.7 atm The downstream pressure, Pd, under standard conditions is 0.01 atm.

Pa = 0.5 x (Pmix + Pd) Pa = 0.505 atm From ANSI N14.5-2014 [Ref. 3]:

MHe = 4.0 g/mol Mair = 29.0 g/mol He = 0.0198 cP air = 0.0185 cP The mass of the mixture of air/helium gases is then determined:

MHe PHe + Mair Pair Mmix = Mmix = 11.5 g/mol Eqn. B.7 from ANSI N14.5-2014 [Ref. 3]

Pmix He PHe + air Pair mix = mix = 0.0194 cP Eqn B.8 from ANSI N14.5-2014 [Ref. 3]

Pmix Change in viscosity as a function of temperature was taken into consideration by using the values listed in Table 4.3.1-1, and performing linear interpolation. Mixture viscosity was determined for each temperature using the same methodology described above.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.3.1-1 Helium and Air Viscosity Temperature Helium Temperature Air Viscosity (Kelvin) Viscosity (cP) (Kelvin) (cP) 250 0.0178 2 273.15 0.0171 4 275 0.0191 2 278.15 0.0173 4 300 0.0201 3 283.15 0.0176 4 350 0.0223 3 288.15 0.0180 4 293.15 0.0182 4 298.15 0.0185 4 303.15 0.0186 4 313.15 0.0187 4 323.15 0.0195 4 333.15 0.0197 4 Determine Lmix as a function of temperature Temperature range for test = T = 273 to 328 K, or equivalently 31.73 °F to 130.73 °F 2.49106 (Dmax )4 Fc(Dmax) = Equation B.3 from ANSI N14.5-2014 [Ref. 3]

amix T

3.81103 (Dmax )3 M mix Fm(T) = Equation B.4 from ANSI N14.5-2014 [Ref. 3]

aPa_

Pa Lmix(T) = (Fc + Fm(T)) (Pmix - Pd) P Equation B.5 from ANSI N14.5-2014 [Ref. 3]

mix Convert the test temperature to Fahrenheit: TF(T) = [(9/5)TK-459.67] °F Figure 4.3.1-1 illustrates the air and helium mixture test leakage rates, Lmix, as a function of temperature in degrees Fahrenheit for helium partial pressures of 0.25, 0.35, 0.45, 0.55, 0.65, 0.75, 0.85, and 0.95 atm.

2 Viscosity based on Reference 15 3

Viscosity based on Reference 13.

4 Viscosity based on Reference 14.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 He 0.95atm 1.90E-07 He 0.85atm 1.70E-07 He 0.75atm 1.50E-07 He 0.7atm Lmix(T) 1.30E-07 He 0.65atm

[ref-cm3/sec] 1.10E-07 He 0.55atm 9.00E-08 He 0.45atm He 7.00E-08 0.35atm He 5.00E-08 0.25atm 20 40 60 80 100 120 140 T [°F]

Figure 4.3.1-1 Allowable Air/Helium Mixture Test Leakage Rates Robatel Technologies, LLC Page 4-12

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The helium component of this leak rate is determined by multiplying the leak rate of the mixture by the ratio of the helium partial pressure to the total mix pressure.

PHe LHe(T) = Lmix(T)*

Pmix 2.50E-07 He 1.00atm He 0.95atm 2.00E-07 He 0.85atm He 1.50E-07 0.75atm LHe (T) He 0.65atm

[ref-cm3/sec] He 1.00E-07 0.55atm He 0.45atm He 5.00E-08 0.35atm He 0.25atm 0.00E+00 20 40 60 80 100 120 140 T [°F]

Figure 4.3.1-2 Allowable Helium Test Leakage Rates Robatel Technologies, LLC Page 4-13

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.3.1-2 Allowable Helium Test Leakage Rates, cm³/sec Helium Helium Helium Helium Helium Leakage Leakage Leakage Leakage Leakage Rate Temperature Temperature Rate PHe- Rate PHe- Rate PHe- Rate PHe-PHe-0.65atm

(°F) (Kelvin) 1.0atm 0.85atm 0.45atm 0.25atm 31.73 273 1.897E-07 1.263E-07 8.185E-08 5.137E-08 2.672E-08 33.53 274 1.898E-07 1.263E-07 8.188E-08 5.138E-08 2.673E-08 35.33 275 1.900E-07 1.264E-07 8.191E-08 5.139E-08 2.673E-08 37.13 276 1.902E-07 1.265E-07 8.194E-08 5.141E-08 2.673E-08 38.93 277 1.903E-07 1.266E-07 8.198E-08 5.142E-08 2.673E-08 40.73 278 1.905E-07 1.267E-07 8.202E-08 5.144E-08 2.674E-08 42.53 279 1.907E-07 1.267E-07 8.205E-08 5.144E-08 2.673E-08 44.33 280 1.909E-07 1.268E-07 8.208E-08 5.144E-08 2.673E-08 46.13 281 1.911E-07 1.269E-07 8.211E-08 5.145E-08 2.672E-08 47.93 282 1.913E-07 1.270E-07 8.214E-08 5.145E-08 2.672E-08 49.73 283 1.914E-07 1.271E-07 8.217E-08 5.146E-08 2.671E-08 51.53 284 1.916E-07 1.272E-07 8.219E-08 5.145E-08 2.670E-08 53.33 285 1.918E-07 1.272E-07 8.221E-08 5.144E-08 2.668E-08 55.13 286 1.920E-07 1.273E-07 8.222E-08 5.144E-08 2.667E-08 56.93 287 1.922E-07 1.274E-07 8.224E-08 5.143E-08 2.666E-08 58.73 288 1.924E-07 1.275E-07 8.226E-08 5.142E-08 2.664E-08 60.53 289 1.925E-07 1.276E-07 8.230E-08 5.144E-08 2.665E-08 62.33 290 1.927E-07 1.277E-07 8.234E-08 5.145E-08 2.665E-08 64.13 291 1.929E-07 1.278E-07 8.238E-08 5.147E-08 2.666E-08 65.93 292 1.931E-07 1.278E-07 8.242E-08 5.148E-08 2.666E-08 67.73 293 1.933E-07 1.279E-07 8.246E-08 5.150E-08 2.667E-08 69.53 294 1.935E-07 1.280E-07 8.249E-08 5.151E-08 2.666E-08 71.33 295 1.936E-07 1.281E-07 8.253E-08 5.151E-08 2.666E-08 73.13 296 1.938E-07 1.282E-07 8.256E-08 5.152E-08 2.666E-08 74.93 297 1.940E-07 1.283E-07 8.259E-08 5.153E-08 2.665E-08 76.73 298 1.942E-07 1.284E-07 8.262E-08 5.153E-08 2.665E-08 78.53 299 1.944E-07 1.285E-07 8.267E-08 5.156E-08 2.666E-08 80.33 300 1.945E-07 1.286E-07 8.272E-08 5.159E-08 2.667E-08 82.13 301 1.947E-07 1.286E-07 8.276E-08 5.161E-08 2.669E-08 83.93 302 1.949E-07 1.287E-07 8.281E-08 5.164E-08 2.670E-08 85.73 303 1.951E-07 1.288E-07 8.286E-08 5.166E-08 2.671E-08 87.53 304 1.952E-07 1.289E-07 8.291E-08 5.169E-08 2.673E-08 89.33 305 1.954E-07 1.290E-07 8.296E-08 5.172E-08 2.675E-08 91.13 306 1.956E-07 1.291E-07 8.301E-08 5.176E-08 2.676E-08 92.93 307 1.958E-07 1.292E-07 8.306E-08 5.179E-08 2.678E-08 94.73 308 1.959E-07 1.293E-07 8.311E-08 5.182E-08 2.680E-08 96.53 309 1.961E-07 1.294E-07 8.317E-08 5.185E-08 2.681E-08 Robatel Technologies, LLC Page 4-14

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.3.1-2 Allowable Helium Test Leakage Rates, cm³/sec (Continued)

Helium Helium Helium Helium Helium Leakage Leakage Leakage Leakage Leakage Rate Temperature Temperature Rate PHe- Rate PHe- Rate PHe- Rate PHe-PHe-0.65atm

(°F) (Kelvin) 1.0atm 0.85atm 0.45atm 0.25atm 98.33 310 1.963E-07 1.295E-07 8.322E-08 5.188E-08 2.683E-08 100.13 311 1.965E-07 1.296E-07 8.327E-08 5.191E-08 2.685E-08 101.93 312 1.966E-07 1.297E-07 8.332E-08 5.194E-08 2.686E-08 103.73 313 1.968E-07 1.297E-07 8.337E-08 5.197E-08 2.688E-08 105.53 314 1.970E-07 1.298E-07 8.340E-08 5.197E-08 2.687E-08 107.33 315 1.971E-07 1.299E-07 8.342E-08 5.197E-08 2.686E-08 109.13 316 1.973E-07 1.300E-07 8.344E-08 5.196E-08 2.685E-08 110.93 317 1.975E-07 1.300E-07 8.346E-08 5.196E-08 2.684E-08 112.73 318 1.977E-07 1.301E-07 8.348E-08 5.196E-08 2.683E-08 114.53 319 1.978E-07 1.302E-07 8.350E-08 5.195E-08 2.682E-08 116.33 320 1.980E-07 1.303E-07 8.352E-08 5.195E-08 2.681E-08 118.13 321 1.982E-07 1.304E-07 8.354E-08 5.195E-08 2.680E-08 119.93 322 1.984E-07 1.304E-07 8.356E-08 5.195E-08 2.679E-08 121.73 323 1.985E-07 1.305E-07 8.359E-08 5.195E-08 2.678E-08 123.53 324 1.987E-07 1.306E-07 8.363E-08 5.197E-08 2.679E-08 125.33 325 1.989E-07 1.307E-07 8.368E-08 5.199E-08 2.681E-08 127.13 326 1.990E-07 1.308E-07 8.373E-08 5.202E-08 2.682E-08 128.93 327 1.992E-07 1.309E-07 8.377E-08 5.205E-08 2.683E-08 130.73 328 1.994E-07 1.310E-07 8.382E-08 5.207E-08 2.685E-08 Figure 4.3.1-2 provides acceptable helium leakage rates at partial helium pressures of 0.25, 0.35, 0.45, 0.55, 0.65, 0.75, 0.85, 0.95, and 1.00 atm. Table 4.3.1-2 provides acceptable helium leakage rates for several helium partial pressures at temperatures ranging from 31.73 °F (273 K) to 130.73

°F (328 K). Figure 4.3.1-2 and Table 4.3.1-2 are to be used to determine the allowable leak rate LHe for the maintenance, fabrication, and periodic leak tests of the RT-100 based on partial pressure of helium and ambient temperatures used in the test. If the measured leakage rate is below the value shown in Figure 4.3.1-2, then the leaktight criteria has been met.

4.3.2 Determination of Equivalent Reference Leakage Rate for Air For the pre-shipment leakage test described in Table 4.1.5-1, the acceptance criteria is based on standard leakage test conditions. NUREG/CR-6487 Section 2.2.6 defines the standard leak rate as corresponding to the upstream volumetric flow rate of dry air with an upstream pressure of 1.0 atmosphere, a downstream pressure of 0.01 atmospheres, and a temperature of 298 K. Tests may be performed at other conditions, provided the acceptance criterion at the testing conditions correspond to the calculated standard leakage rate acceptance criterion [Ref. 4]. The method for determining the corresponding leak rate is described in ANSI N14.5-2014 Section B.4.4 [Ref. 3].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Two pre-shipment leak test procedures are described in Chapter 8 of this SAR; a gas-pressure rise method (Section 8.2.2.2), and a gas-pressure drop method (Section 8.2.2.3). The gas-pressure drop method requires a set of conditions different than the standard leak rate conditions. In order to simplify this leak test, a pressure differential was selected that corresponds to a sensitivity of 1x10-3 cm3/sec. Given atmospheric air pressure conditions of 1 atm and a temperature of 298K, the upstream air pressure should be 1.67 atm, as described in Calculation Package 2014-020-CALC-LT-001, Rev. 0 [Ref. 25].

4.4 Hydrogen Gas Generation Hydrogen gas buildup in loads containing waste material typically occurs due to radiolysis of hydrogenous material in the contents. As hydrogen is generated, it could potentially accumulate within the cask cavity in flammable concentrations. Based on USNRC guidance, the flammability limit of 0.05 volume fraction (mole fraction and volume fraction is interchangeable when discussing ideal gas buildup) hydrogen in air was measured in accordance with NUREG/CR-6673 Hydrogen Generation in TRU Waste Transportation Packages [Ref. 16], and supplemented with data from EPRI NP-5977 Radwaste Radiolytic Gas Generation Literature Review [Ref.

21].

Materials that make up the contents that can undergo radiolysis include primarily ion exchange resins, with lesser quantities of polystyrene and polyamides (nylon). Materials that make up the secondary container and shoring that can undergo radiolysis include polyethylene, wood, and polypropylene. Free water in the contents and moisture in the resin beads are also included in the analysis. In order to provide a bounding analysis, it is assumed that all of the decay energy in the contents produces gas generation in the waste or secondary container.

The rate of gas generation by radiolysis in these materials is dependent upon the type of incident radiation. Alpha emitters tend to generate more hydrogen per unit of energy deposited than gamma/beta emitters. Typical resin and filter waste produced at commercial nuclear power reactor facilities contains a high percentage of gamma in relation to alpha emitters. Typical examples of historical shipment data are provided with RT100-REF-01-01 Historical Cask Summaries by Waste Category [Ref. 22]. Because NUREG/CR-6673 is primarily focused on the alpha radiation predominant in TRU waste, EPRI NP-5977 is utilized to obtain gamma radiation G Values for these primary waste materials.

The typical shipment data referenced above indicates that the decay energy is approximately 90~100% from gamma radiation. In order to bound these shipments and to facilitate the utilization of a loading curve as a function of decay heat and waste volume in the cask, a decay energy distribution of 80% gamma and 20% alpha decay energy is assumed. The evaluation presented in Sections 4.4.1, 4.4.2, and 4.4.3 utilize this distribution as a way of illustrating the calculational method. Section 4.4.4 provides the user with a simplified model used to develop the loading curve Robatel Technologies, LLC Page 4-16

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 (Figure 4.4.4-1) in order to determine the maximum allowable decay heat as a function of waste volume. Section 4.4.5 provides the user with an analytical model along with a set of G Values for the bounding waste material as a function of the gamma and alpha decay energy distribution for cases that do not fit the loading curve.

Package material and content that can generate flammable gas shall be appropriately assigned as part of the ionic resin bead waste or polyethylene container when using the Loading Curve (Figure 4.4.4-1) or detailed analysis (Section 4.4.5) to determine acceptable hydrogen gas generation-related parameters of shipping time and decay heat. For example, waste filters (made of material other than polypropylene or polyethylene) shall be grouped as ionic bead waste and wood shoring would be grouped as part of the polyethylene container. If filters are made of polyethylene or polypropylene, they are to be included in the secondary container volume for the hydrogen gas generation detailed analysis. When no hydrogenous materials are included in the contents, a hydrogen buildup calculation is not required. For example, if the only radioactive contents are activated hardware and the secondary container and all shoring are metallic or non-hydrogenous, no calculations for hydrogen gas generation and buildup are required. When mixed with hydrogen generating contents, the effect of non-hydrogenous materials, such as activated hardware contents, is only from the volume that they occupy in the cavity.

4.4.1 Determination of Bounding G Values The first step in performing a gas generation calculation is to determine the G Values. As such, the following sections describe the steps in this process.

4.4.1.1 G Values for Waste and Secondary Container Materials A list of G Values is provided in Table 4.4.1-1 and are taken from NUREG/CR-6673 [Ref. 16],

EPRI NP-5977 [Ref. 21], and RH-TRU 72-B SAR [Ref. 23]. These materials represent all potential cask contents as indicated in Section 1.2.2.3, Physical and Chemical Form - Density, Moisture Content and Moderators. Potential materials in the waste that can undergo radiolysis are polystyrene, nylon, polyamides, ion exchange resins, and any residual water. Secondary container and shoring materials include polyethylene, wood and polypropylene.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.1-1 G Values (Molecules/100eV) for Potential Content Materials GH GFG GT Material G (H2) G (flammable gas) G (net gas)

Waste Materials Polystyrene (Alpha Radiation) 0.20 0.20 0.20 Polyamides/Nylon5 1.10 1.20 1.50 Ion Exchange Resins (Alpha Radiation) 1.70 1.70 2.10 Ion Exchange Resins (Gamma/Beta 0.62 0.62 0.62 Radiation)6 Water (Liquid Phase, Gamma 0.45 0.45 0.45(8)

Radiation)7 Water (Liquid Phase, Alpha Radiation) 1.60 1.60 1.60(8)

Paper 0.90 0.90 1.50 Polyethylene Filter 4.00 4.10 4.10 Polypropylene Filter 3.30 3.40 3.40 Secondary Container / Shoring Polyethylene 4.00 4.10 4.10 Wood (Cellulose) 3.20 3.20 10.20 Polypropylene 3.30 3.40 3.40 The ion exchange resin has the highest flammable gas G Value due to alpha radiation when compared to the other hydrogenous materials that could be contained within the waste. The G Value of ionic resin for gamma radiation is taken from EPRI NP-5977, which indicates that fully swollen ionic exchange resins have flammable gas G Values of up to 0.62.

Ionic resins are dewatered before transport, meaning most of the free water is removed. Even in a fully dewatered state, ion exchange resin beads can contain from 50% to 66% moisture, per NUREG/CR-6673 and EPRI NP-5977. The term dewatered should not be confused with the term dry for ionic resins. Based on Section 4 of EPRI NP-5977, the G Values for fully dried resins are a factor of 10 less than swollen resin beads (from 0.001 to 0.067 in recorded experiments). As such, it can be concluded that the G values for ionic resins are primarily driven by moisture content and the values utilized in Table 4.4.1-1 already take into consideration the moisture content in the resin.

5 Based on NUREG/CR-6673, Section D.7.22 [Ref. 16], nylon is a polyamide. Polyamides are bounded by these values.

6 The GFG value for ionic resin is used for GT because no value is provided in EPRI NP-5977 [Ref. 21]. Less non-flammable gas production will decrease the amount of time required to achieve a flammable mixture, making this a bounding assumption.

7 Based on NUREG/CR-6673 Table D.1 [Ref. 16], the largest G(H2) for liquid water subjected to gamma radiation is 0.45 molecules/100eV.

8 For water, The GT value is set to the GFG value, as explained later in Section 4.4.1.1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Only hydrogen gas was considered as a byproduct of the radiolysis of water. This results in the fraction of flammable gas to the total gas generated () of 1.0 in Equation 4.8 of NUREG/CR-6673 [Ref. 16]. Including oxygen in the total gas generation from the radiolysis of water would decrease the mole fraction of hydrogen (XH) in the free gas volume. This is because the alpha term would be less than 1.0. Thus, using the value of 1.0 would yield the most bounding result.

4.4.1.2 Calculation of Effective G Values Table 4.4.1-1 lists the G Values for the hydrogenous material that could be transported in the cask.

Both alpha and gamma G Values are provided for the most predominant waste contents of resin and water. For other materials, the more conservative alpha radiation values are utilized. As noted in Section 4.4, hydrogen gas generation calculations for typical resin waste contents are performed assuming that the decay energy of the waste is 80% gamma and 20% alpha. The effective G Values for these materials is calculated using these fractions applied to the corresponding G Value.

Materials without a gamma G Value are taken as the alpha G Value. The effective G Values are provided in Table 4.4.1-2.

Table 4.4.1-2 Effective G Values (Molecules/100eV) for Potential Content Materials Effective GH Effective GFG Effective GT Material G (H2) G (flam. gas) G (net gas)

Waste Materials Polystyrene 0.20 0.20 0.20 Polyamides/Nylon 1.10 1.20 1.50 Ion Exchange Resins 0.84 0.84 0.92 Water 0.68 0.68 0.68 Paper 0.90 0.90 1.50 Polyethylene Filter 4.00 4.10 4.10 Polypropylene Filter 3.30 3.40 3.40 Secondary Container / Shoring Polyethylene 4.00 4.10 4.10 Wood (Cellulose) 3.20 3.20 10.20 Polypropylene 3.30 3.40 3.40 Robatel Technologies, LLC Page 4-19

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.4.1.3 Operating Temperature G Value Adjustment As described in Section 2.4.2 of NUREG/CR-6673 [Ref. 16], the hydrogen gas generation rate of some materials is noticeably affected by the temperature in the container during transport. This is contingent upon the activation energy of the material being shipped in the cask. The activation energies for the materials used in the hydrogen generation calculations are shown in Table 4.4.1-3, and are based on Table 3.11 of NUREG/CR-6673 and RH-TRU 72-B Appendices [Ref. 23]. The activation energy for ionic resin is not specifically listed in NUREG/CR-6673, but RH-TRU 72-B SAR specifies that organic resins have an activation energy of 2.1 kcal/mole.

Table 4.4.1-3 Activation Energy Activation Energy Material (kcal/mole)

Waste Materials Polystyrene 0.8 Polyamides/Nylon 0.8 Resins 2.1 Water 0.0 Paper 1.3 Polyethylene Filter 0.8 Polypropylene Filter 0.8 Secondary Container / Shoring Polyethylene 0.8 Wood 2.1 Polypropylene 0.8 The G value at NCT temperatures is determined using Equation 2.2 of NUREG/CR-6673.

2 1 2 = 1 exp [( ) ( )]

2 1 where: GT1 = radiolytic G value at 298 K [molecules/100eV]

GT2 = radiolytic G value at transport temperature [molecules/100eV],

Ea = activation energy for radiolytic gas generation [kcal/gmol],

R = gas law constant [1.987x10-3 kcal/gmol-K],

T1 = 298 K T2 = temperature of contents during transport [K]

Based on Table 3.1.3-1, the maximum inner shell temperature during NCT is 73.1 °C. Therefore, a bounding value of 80 °C (353.15 K) is utilized for the contents in this analysis. For example, the resultant GFG value (and GT value because they are the same for polyethylene) for polyethylene at Robatel Technologies, LLC Page 4-20

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 353.15 K is equivalent to:

0.8/ 353.15 298

= (4.1/100)exp [( 3

)( )]

1.98710 / (353.15)(298)

= 5.06 molecules/100eV The final G values used in the hydrogen generation calculations are shown in Table 4.4.1-4.

Table 4.4.1-4 Bounding G Values for Contents at Maximum NCT Temperature Material G (H2), GH G (flammable G (net gas), GT gas), GFG Waste Materials Polystyrene 0.25 0.25 0.25 Polyamides 1.36 1.48 1.85 Ion Exchange Resins 1.45 1.45 1.59 Water 0.68 0.68 0.68 Paper 1.27 1.27 2.11 Polyethylene Filter 4.94 5.06 5.06 Polypropylene Filter 4.08 4.20 4.20 Secondary Container / Shoring Polyethylene 4.94 5.06 5.06 Wood 5.57 5.57 17.75 Polypropylene 4.08 4.20 4.20 Of the materials that could comprise the waste, resin and water are present in the greatest quantities. While polyamides have a slightly higher G(flammable gas) value than the resins, resins were chosen as the bounding contents because resins have a much higher density when loaded than the polyamides which form a small part of filters. In addition, hydrolysis of polyamides produces nonflammable gas which would tend to dilute hydrogen concentration. If polyethylene or polypropylene filters are loaded into the cask, their volumes shall be accounted for as a polyethylene secondary container in the calculation. Therefore, resin and water are selected for utilization in the gas generation calculations.

The secondary container and shoring materials are assumed to be polyethylene. Like polyamides in the waste, wood has a slightly higher G Value than polyethylene. However, wood has a significantly higher total gas G Value, which offsets the impact of flammable gas generation by generating more than 2 moles of non-flammable gas for every mole of flammable gas.

Additionally, the wood would be present only in limited quantities as shoring material on the outside of the secondary container.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 As described in Section 4.4, the gas generation analysis is performed assuming that all decay energy is absorbed in either the waste or the secondary container, maximizing the amount of gas generated through radiolysis. In fact, much of the gamma radiation emitted from the waste escapes the cavity and is absorbed in the casks lead shielding material.

These G values are then utilized to calculate the hydrogen gas generation rates as described in Section 4.4.3.

4.4.2 Hydrogen Gas Generation by Radiolysis For the hydrogen generation evaluation, the RT-100 is treated as a single rigid non-leaking enclosure. Using Equation 4.8 on page 31 of NUREG/CR-6673 [Ref. 16], an equation characterizing the mole fraction of hydrogen (or flammable gas) in the RT-100 over time for a single material generating hydrogen is shown below.

Equation 4.4 100

0 + 0

+

0 100 where: XH = mole fraction of hydrogen, nH = number of moles of hydrogen [gmol],

n0 = initial number of gas moles in the container when the vessel was closed

[gmol],

nnet = number of moles of gas generated [gmol],

GT = total radiolytic G value [molecules/100eV],

DH = decay heat that is absorbed by the radiolytic materials [eV/s],

= fraction of GT that is equivalent to GFG, flammable gas released, AN = Avogadros constant [6.022 x 1023 molecules/gmol],

P0 = pressure when the container is sealed [atm],

T0 = temperature when the container is sealed [K],

V = is the container void volume [cm3],

Rg = gas law constant [82.05 cm3atm/gmolK],

t = time [seconds]

Based on Section 5 of NUREG/CR-6673 [Ref. 16], shipping periods other than one year need to be defined as one half the time it takes for hydrogen to accumulate in the package to a concentration equivalent to the lower flammability limit. To ensure that this is taken into consideration in the calculations, the equation above has been adjusted to incorporate a multiple of 2 times the shipping period required. Equation 4.4 is also limited to providing hydrogen mole fraction over time for one hydrogenous material. In this analysis there are three hydrogenous materials that are taken into consideration, water in the waste material, the resin, and the polyethylene container. The resultant Robatel Technologies, LLC Page 4-22

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 equation generated once these parameters are taken into consideration (the increase in shipping time and the number of hydrogenous materials) is shown below in Equation 4.5.

Equation 4.5 (2) (2) (2)

+ 100 + 100 100

=

0 (2) (2) (2) 0 100 + 100 + 100

+

(2) (2) (2)

+ 100 + 100 100

=

0 (2) (2) (2) 0 + 100 + 100 + 100 where: XH = mole fraction of hydrogen, GTi = total radiolytic G value for ionic resin and stainless steel filters

[molecules/100eV],

GTC = total radiolytic G value for polyethylene container, shoring, and polyethylene or polypropylene filters [molecules/100eV],

GTW = total radiolytic G value for water in waste [molecules/100eV],

DH = decay heat that is absorbed by the radiolytic materials [eV/s],

Di = decay heat that is absorbed by the ionic resin and stainless steel filters [eV/s],

DC = decay heat that is absorbed by the polyethylene container, shoring, and polyethylene or polypropylene filters [eV/s],

DW = decay heat that is absorbed by the water [eV/s],

i = fraction of GTi that is equivalent to GFGi, flammable gas released, for the ionic resin and stainless steel filters, C = fraction of GTC that is equivalent to GFGC, flammable gas released, for the secondary container, shoring, and polyethylene or polypropylene filters in the waste, W = fraction of GTW that is equivalent to GFGW, flammable gas released, for the water in waste, FW = fraction of decay heat energy absorbed by the water in the waste material, Fi = fraction of decay heat energy absorbed by the ionic resin and stainless steel filters in the waste material, FC = fraction of decay heat energy absorbed by the polyethylene container, shoring, and polyethylene or polypropylene filters, AN = Avogadros constant [6.022 x 1023 molecules/gmol],

P0 = pressure when the container is sealed [atm],

T0 = temperature when the container is sealed [K],

V = is the container void volume [cm3],

Rg = gas law constant [82.05 cm3atm/gmol-K],

t = time [seconds]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.4.3 Hydrogen Generation - Radiolysis in Waste, Water and Polyethylene Container In order to determine the time available to transport the RT-100, Equation 4.5 must be manipulated to provide time limit versus waste volume and decay heat of inventory. Given the decay heat and other cask content parameters, the time to reach 5% by volume of combustible gases is determined as follows.

Solve for t, (2) (2) (2)

+ 100 + 100 100

=

0 (2) (2) (2) 0 + 100 + 100 + 100 0 (2) (2) (2)

( + + + )=

0 100 100 100 (2) (2) (2)

+ +

100 100 100 50 0 0

=

[ ] + [ ] + [ ]

Alternatively, given the limiting transport time and other cask parameters, the Equation 4.5 must be manipulated to provide decay heat limit versus the waste volume and the shipping period. The decay heat limit versus the free gas volume and shipping period (all decay heat energy deposited into the waste material and the polyethylene container) is determined as follows.

Solve for DH, (2) (2) (2)

+ 100 + 100 100

=

0 (2) (2) (2) 0 + 100 + 100 + 100 0 (2) (2) (2)

( + + + )=

0 100 100 100 (2) (2) (2)

+ +

100 100 100 Robatel Technologies, LLC Page 4-24

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 50 0 0

=

[ ] + [ ] + [ ]

where: XH = mole fraction of hydrogen, GTi = total radiolytic G value for ionic resin and stainless steel filters

[molecules/100eV],

GTC = total radiolytic G value for polyethylene container, shoring, and polyethylene or polypropylene filters [molecules/100eV],

GTW = total radiolytic G value for water in waste [molecules/100eV],

DH = decay heat that is absorbed by the radiolytic materials [eV/s],

Di = decay heat that is absorbed by the ionic resin and stainless steel filters [eV/s],

DC = decay heat that is absorbed by the polyethylene container, shoring, and polyethylene or polypropylene filters [eV/s],

DW = decay heat that is absorbed by the water [eV/s],

i = fraction of GTi that is equivalent to GFGi, flammable gas released, for the ionic resin and stainless steel filters, C = fraction of GTC that is equivalent to GFGC, flammable gas released, for the secondary container, shoring, and polyethylene or polypropylene filters in the waste, W = fraction of GTW that is equivalent to GFGW, flammable gas released, for the water in waste, FW = fraction of decay heat energy absorbed by the water in the waste material, Fi = fraction of decay heat energy absorbed by the ionic resin and stainless steel filters in the waste material, FC = fraction of decay heat energy absorbed by the polyethylene container, shoring, and polyethylene of polypropylene filters, AN = Avogadros constant [6.022 x 1023 molecules/gmol],

P0 = pressure when the container is sealed [atm],

T0 = temperature when the container is sealed [K],

V = is the container void volume [cm3],

Rg = gas law constant [82.05 cm3atm/gmolK],

t = time [seconds]

The next step is to determine the values of the variables in this equation. Avogadros constant and the gas law constant are known values set at 6.022 x 1023 molecules/gmol and 82.05 cm3atm/gmolK, respectively, in this analysis. Initial gas temperature and pressure have been set at maximum NCT conditions (311 K and 1 atm). Based on USNRC guidance, the flammability limit of 0.05 volume fraction hydrogen in air was measured NUREG/CR-6673 [Ref. 16]. The radiolytic G values and values utilized are as provided in Table 4.4.3-1. The time required has been arbitrarily set at 10 days (864,000 seconds) in this analysis (reminder that equations automatically double the time entered into the equation based on guidance suggested in NUREG/CR-6673 [Ref.

16]).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.3-1 Effective G Values and Corresponding Values for Contents Gamma G (flam G (net gas),

Alpha Frac. Material Frac. gas), GFG GT 80% 20% Polyethylene 5.06 5.06 1.00 80% 20% Resin 1.45 1.59 0.91 80% 20% Water 0.68 0.68 1.00(9)

Determining the fraction of decay heat energy absorbed by the free water in the waste material (FW), ionic resin and stainless steel filters in the waste material (Fi), and by the polyethylene container, shoring, and polyethylene or polypropylene filters in the waste material (FC) is approximated by assuming that the energy absorbed is proportional to the volume of the material in question divided by the total volume of hydrogenous material being shipped in the cask. It is assumed the decay heat is entirely absorbed by hydrogenous materials and none is absorbed or dissipated by the activated hardware contents which is conservative. Thus, the fractions can be described as follows:

=

+ +

=

+ +

=

+ +

where: VW = volume of free water in the waste Vi = volume of dewatered10, ionic resin in the waste, including absorbed moisture, and stainless steel filters in the waste material VC = volume occupied by the secondary container, shoring, and polyethylene or polypropylene filters in the waste material Water is present in resin bead shipments in two distinct forms; one is absorbed moisture within the resin bead itself; the other is free water that is present between the resin beads. The absorbed moisture within the resin bead is considered in the hydrogen generation analysis as it is incorporated into G Value of the resins.

9 For water, The GT value is set to the GFG value to obtain an value of 1.0.

10 The term dewatered resin refers to resins in which free water has been removed from between the resin beads at the time of preparation for storage or transportation. The amount of free water in dewatered resins is typically around 1% after mechanical draining (EPRI NP-5977 Radwaste Radiolytic Gas Generation Literature Review, page 10 [Ref. 21]). The term dewatered should not be confused with dry. Dewatered resin beads shipped in the RT-100 could have a moisture content up to 50~66% based on NUREG/CR-4062 Extended Storage of Low-Level Radioactive Waste, Potential Problem Areas [Ref. 20], and EPRI NP-5977.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The volume occupied by the waste (VWASTE) is assumed to be 99% ionic resin and 1% free water.

The ionic resin beads are assumed to be uniform spheres. Since the random packing fraction for uniform spheres is 0.64 [Ref. 5], the remaining free space (0.36) of the waste volume is assumed to be 0.11 air and 0.25 water to account for grossly dewatered material content. Thus, the volumes of ionic resin and water in the waste can be described as follows:

= (0.99)(0.25) + (0.01) = (0.2575)

= (0.99)(0.64) = (0.6336)

As noted, the assumed content is either dewatered resin or grossly dewatered resin. The amount of free water in dewatered resins10 is typically around 1% after mechanical draining, by regulation no more than 1% to meet disposal requirements. Grossly dewatered resins have a higher free water amount that shall be limited to 20% of the ionic resin volume (20.75% of the waste volume).

Therefore, the amount of free water assumed (25.75% of waste volume) thus represents a free water volume that bounds the dewatered resin state by an order of magnitude, and bounds the grossly dewatered resin state by around 25%, and was chosen to represent a bounding condition for hydrogen generation.

Therefore, the fraction of decay heat energy absorbed by the water in the waste volume (FW), ionic resin and stainless steel filters in the waste volume (Fi), and by the polyethylene container, shoring, and polyethylene or polypropylene filter (FC) are equivalent to:

0.2575

=

(0.2575 + 0.6336 + )

0.6336

=

(0.2575 + 0.6336 + )

=

(0.2575 + 0.6336 + )

The remaining 0.11 fraction of ionic resin volume is air. The remaining fraction of air volume in the ionic resin is taken into consideration in the total free gas volume (V). Where the free gas volume (V) is equivalent to the total cavity volume (4.60x106 cm3) minus the sum of the container, shoring, and polyethylene or polypropylene filter volume (VC), water volume (VW), the ionic resin and stainless steel filter volume (Vi), and the volume of any activated hardware contents (VH), as applicable.

= (4.60 x 106 3 ) ( + + + )

= (4.60 x 106 3 ) ( + 0.2575 + 0.6336 + )

= (4.60 x 106 3 ) 0.8911 Robatel Technologies, LLC Page 4-27

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Incorporating the equations for fraction of decay heat energy absorbed by the water in the waste volume, ionic resin in the waste volume and by the polyethylene container into the derivation of time limit and decay heat limit results in the following equations.

Equation 4.6 50 0 (0.2575 + 0.6336 + )

0

=

(0.6336 ) [ ] + ( ) [ ] + (0.2575 ) [ ]

Equation 4.7 50 0 (0.2575 + 0.6336 + )

0

=

(0.6336 ) [ ] + [ ] + (0.2575 ) [ ]

The final step to solving the equation is determining the free gas volume which will vary based on the inventory in the RT-100. The maximum cavity volume is known to be 4.60x106 cm3 (162.37 ft3) based on Table 4.1.5-2. In order to determine the free gas volume an approximation of the volume occupied by the polyethylene liner needs to be made. The guiding technical issue in determining the free gas volume is to maximize the hydrogen gas mole fraction buildup rate that then results in a conservative shipping time (limiting the allowable shipping time). A greater hydrogen mole fraction buildup rate in the cavity is produced by minimizing the free gas volume of the cavity. Minimizing the available free gas volume is accomplished by using the polyethylene container with the largest container volume. Exhibit A of Cask Procurement Agreement dated April 10, 2012 [Ref. 17], provides the burial volume, maximum internal volume, and empty weight of various containers, and is shown in Table 4.4.3-2.

The container volume may be calculated from the empty weight and material density. In order to calculate the largest container volume, a minimum material density is used. The material density of high density polyethylene is 0.959 g/cm3, while the material density of plain carbon steel is 7.85 g/cm3 [Ref. 24]. For a bounding assumption, and to take into account empty liner weight tolerances, the densities were reduced by 10%, resulting in a density of 0.863 g/cm3 (53.88 lb/ft3) for polyethylene and 7.065 g/cm3 (441.05 lb/ft3) for steel. The result is that the EL-142 container has the largest container volume of 27.02 ft3. A volume of 30.1 ft3 was used in the analysis to represent the volume occupied by the secondary container and any shoring. Therefore, the maximum free gas volume (no waste volume) is 132.27 ft3. If a different container is used in the RT-100 transport cask, the equations generated in this section of the SAR can be adjusted.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The analysis assumes use of a polyethylene container. As described in Section 4.4.1.3 this is considered a bounding assumption based on material G Values. Use of a steel liner listed in Table 4.4.3-2 is considered acceptable because steel does not contribute to hydrogen gas generation.

Given the 10 day limiting transport time, the concern is how much waste volume and decay heat is acceptable for the individual shipments. The waste volume is then equal to the maximum cavity volume subtracted by the free gas volume and the polyethylene container volume.

A loading curve of allowable decay heat as a function of waste volume is provided in Section 4.4.4 for a specific set of waste parameters, including G-Values based on a bounding decay heat distribution of 80% gamma and 20% alpha. Additionally, in case a detailed analysis is performed, a procedure is given in Section 4.4.5, and a list of effective G Values for other distributions of gamma and alpha radiation is provided in Table 4.4.5-2.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.3-2 Secondary Container Volumes and Allowable Shoring Volume Volume Allowable Empty Material Density12 Occupied by Shoring Container Weight11 Type (lb/ft3) Container13 Volume (ft3)

(lbs)

(ft3)

PL 6-80 MT Polyethylene 500 53.88 9.28 20.82 PL 6-80 MTIF Polyethylene 525 53.88 9.74 20.36 PL 6-80 FR Polyethylene 550 53.88 10.21 19.89 PL 6-80 Polyethylene 625 53.88 11.60 18.50 FP/FEDX PL 8-120 MT Polyethylene 600 53.88 11.14 18.96 PL 8-120 MTIF Polyethylene 625 53.88 11.60 18.50 PL 8-120 FR Polyethylene 650 53.88 12.06 18.04 PL 8-120 Polyethylene 725 53.88 13.46 16.64 FP/FEDX PL 8-120 CMT Polyethylene 720 53.88 13.36 16.74 PL 14-150 Polyethylene 800 53.88 14.85 15.25 PL 10-160 MT Polyethylene 700 53.88 12.99 17.11 PL 10-160 MTIF Polyethylene 735 53.88 13.64 16.46 PL 10-160 FR Polyethylene 750 53.88 13.92 16.18 PL 10-160 Polyethylene 825 53.88 15.31 14.79 FP/FEDX NUHIC-55 Polyethylene 150 53.88 2.78 27.32 NUHIC-136 Polyethylene 600 53.88 11.14 18.96 Radlok 500 Polyethylene 680 53.88 12.62 17.48 EL-50 Polyethylene 909 53.88 16.87 13.23 EL-142 Polyethylene 1456 53.88 27.02 3.08 L 6-80 MT Steel 1000 441.05 2.27 27.83 L 6-80 CMT Steel 1150 441.05 2.61 27.49 L 6-80 IN-SITU Steel 3500 441.05 7.94 22.16 L 6-80 FP Steel 1050 441.05 2.38 27.72 L 6-80 FP/FEDX Steel 1225 441.05 2.78 27.32 L 8-120 MT Steel 1200 441.05 2.72 27.38 L 8-120 CMT Steel 1350 441.05 3.06 27.04 L 8-120 IN-SITU Steel 4200 441.05 9.52 20.58 L 8-120 FR Steel 1250 441.05 2.83 27.27 L 8-120 Steel 1325 441.05 3.00 27.10 FP/FEDX ES-50 Steel 250 441.05 0.57 29.53 ES-142 Steel 1100 441.05 2.49 27.61 11 From Exhibit A of Cask Procurement Agreement dated April 10, 2012 by and between Waste Control Specialists LLC and Robatel Technologies, LLC et al [Ref. 17].

12 The bounding calculation assumes a maximum container volume. Therefore, lower density values of 0.863 g/cm 3 and 7.065 g/cm3 were chosen for polyethylene and steel.

13 Calculated as the Empty Weight divided by the Density, neglecting void space.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.4.4 Hydrogen Gas Generation - Simplified Model used to develop Loading Curve Using Equation 4.7, the decay heat limit versus waste volume can be determined for a limit of 5%

in the cavity free volume. Figure 4.4.4-1 provides a curve illustrating the waste volume to decay heat value that would result in the generation of a flammable gas mixture within 10 days assuming that all decay heat is absorbed by the waste material and the polyethylene container. The calculation assumes that the hydrogen generation occurs over a period of time that is twice the allowable shipping time. Note that for this simplified model, similar to the filter materials, any activated hardware waste in the contents is grouped as ionic bead waste and is assumed to be hydrogen generating. The non-hydrogen gas generating characteristic of the hardware waste is only credited in the analytical model for the hydrogen gas buildup calculations outlined in Section 4.4.5.

For most shipments, this simplified graphical model (Loading Curve) can be used to determine the maximum heat load. However, use of the Loading Curve is limited to the restrictions noted in Table 4.4.4-1. One restriction of using the Loading Curve is that the secondary container is listed in Table 4.4.3-2, or is a container of equivalent material volume.

If the waste volume and decay heat values for the contents fall below the Loading Curve illustrated in Figure 4.4.4-1, and the restrictions listed in Table 4.4.4-1 are met, the load would not generate a flammable gas mixture during shipment. Otherwise, the user must perform a more detailed calculation of hydrogen generation for their specific contents and expected shipping time using the information provided in Section 4.4.5. The use of this calculation ensures that the requirements of NUREG/CR-6673 [Ref. 16] are met.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 4.4.4-1 Package Loading Curve for Hydrogen Generation - Decay Heat Limit Versus Waste Volume Robatel Technologies, LLC Page 4-32

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.4-1 Conditions and Justifications for using Package Loading Curve (Figure 4.4.4-1)

Condition for Shipper to use Justification Loading Curve Waste consisting of spent resins/filters Historically shipments of commercial resins/filters and activated or mixture of spent resins/filters with hardware have consisted of approximately 90-100% gamma radiation activated hardware from commercial [Ref. 22]. To bound these shipments, the calculation assumes a decay power plants energy distribution of 80% gamma and 20% alpha radiation. This results in effective G-Values for resin beads of 0.836 and for water of 0.68 For mixed shipments, activated molecules/100eV at 298K. These values are adjusted for maximum NCT hardware are grouped with resins and temperature of 80 °C, and bound the expected G-Values for resins and filters when the simplified model is filters from commercial reactors.

used.

Waste has been dewatered or grossly The loading curve assumes a free water volume of 25.75%. The main dewatered effect of free water is to limit the void volume in the cavity, thereby increasing the hydrogen mole fraction. Since the loading curve assumes 25.75% free water for a hydrogen concentration of 5% or less, shippers must ensure that the limit is not exceeded.

No limit on moisture content of resin The G-values for resins, as noted in Table 5.3 of EPRI NP-5977, are for resins with high moisture contents (i.e., swollen resin).

Use of a liner (or equivalent) listed in The calculation determines the free volume for waste by subtracting the Table 4.4.3-2 with maximum shoring maximum liner and shoring volume from the cask cavity volume.

volume as specified Equivalent liners may be used provided the volume occupied by the liner and shoring material does not exceed 30.1 ft3.

Shipment time not greater than 10 days Shipment time calculated for 20 days (allowing a shipment within 10 days following regulation). This is true for resins/filters shipments and mixture of resins/filters with activated hardware shipments when the simplified loading curve is used.

Loading at temperature not to exceed The maximum ambient NCT temperature is 38 °C per 10 CFR 71.

38 °C and standard pressure (1 atm)

Secondary containers are passively Secondary containers are required to be passively vented within the cask vented within the cask cavity during cavity. The loading curve assumes the gases generated are free to occupy shipment. the cask cavity volume inside and outside the secondary container.

Filters are not made of polyethylene or Polyethylene or polypropylene filters have higher G Values than resins, polypropylene therefore, they require performing the detailed analysis and addition of their volume to the secondary container volume (VC).

Waste volume not greater than 130 ft3 Extrapolation beyond the curve is not allowed.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.4.5 Hydrogen Gas Generation - Analytical Model used for Detailed Analysis If Figure 4.4.4-1 is not applicable to a shipment, or if further analysis is required, the equations derived in Section 4.4.3 can be used. With substitution and further simplification of Equation 4.6 and Equation 4.7, the maximum allowable shipping time (tmax) or the maximum allowable decay heat (DH,max) can be described as:

Equation 4.8 (2.5 0 )(4.66 0.8911 )(0.8911 + )

=

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

Equation 4.9 (2.5 0 )(4.66 0.8911 )(0.8911 + )

, =

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

where: AN = Avogadros constant [6.022x1023 molecules/gmol]

Rg = gas law constant [82.05 cm3atm/gmolK]

P0 = pressure when the container is sealed [atm]

T0 = temperature when the container is sealed [K]

DH = decay heat of cask contents [eV/s]

t = shipment time [s]

VC = volume occupied by the secondary container, shoring, and polyethylene or polypropylene filters in the waste [cm3]

VWASTE = volume occupied by the ionic resin and stainless steel filters in the waste material [cm3]

VH = volume occupied by activated hardware in the waste material (combined volumes of low-density and high-density hardware) [cm3]

GTi = total radiolytic G value for the ionic resin and stainless steel filters

[molecules/100eV]

GTC = total radiolytic G value for the secondary container, shoring, and polyethylene or polypropylene filters in the waste [molecules/100eV]

GTW = total radiolytic G value for water in waste [molecules/100eV]

i = fraction of GTi that is equivalent to GFGi, flammable gas released, for the ionic resin and stainless steel filters C = fraction of GTC that is equivalent to GFGC, flammable gas released, for the secondary container, shoring, and polyethylene or polypropylene filters in the waste W = fraction of GTW that is equivalent to GFGW, flammable gas released, for water in the waste Note 1: Use of Equation 4.8 and Equation 4.9 are valid only when the conditions listed in Table 4.4.5-1 are met. Shipments are allowed only if the conditions in Table 4.4.5-1 are met.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.5-1 Conditions for Shipper to use the Detailed Analysis 1 Waste consists of resins/filters or mixture of resins/filters with activated hardware from commercial power plants. Activated hardware only shipments do not require hydrogen gas analysis.

2 Waste has been grossly dewatered.

3 Secondary containers are passively vented within the cask cavity during shipment.

The user may measure the decay heat of the cask contents (DH) in order to calculate the maximum allowable shipping time (tmax) using Equation 4.8.

Alternately, the user may know the shipment time (t) and calculate the maximum allowable decay heat of the cask contents (DH,max) using Equation 4.9.

Initial pressure (P0) and initial temperature (T0) may be measured by the user at the time of loading.

The volume occupied by the secondary container, shoring, and polyethylene or polypropylene filters in the waste (VC), the volume occupied by the ionic resin and stainless steel filters in the waste material (VWASTE), and the activated hardware volume (VH) are known.

The use of different G-values (GTi, GTC, GTW) and fractions (i, C, W) must be justified by the user based on waste characterization. These variables must be adjusted for the transport temperature of 80°C as described in Section 4.4.1.3, in order to meet the requirements of NUREG/CR-6673 [Ref. 16].

The values must also be adjusted for the appropriate alpha/gamma radiation distribution. One example of this adjustment is provided in Table 4.4.5-2 for the same G-values in the bounding case loading curve for the 80% gamma/20% alpha decay heat distribution.

Note 2: Alternatively, the user can follow the NUREG/CR-6673 requirements to determine the shipping time to reach a hydrogen concentration of 5%. The shipping time has to be defined as 1/2 the time to reach the 5% hydrogen concentration per the requirement in NUREG/CR-6673 [Ref.

16].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 4.4.5-2 G-values and -Fractions for a Range of Alpha/Gamma Decay Heat Distributions G (flam gas), G (net gas),

Gamma Frac Alpha Frac Material G (H2), GH GFG GT Polyethylene 4.94 5.06 5.06 1.00 0.0 1.0 Resin 2.96 2.96 3.65 0.81 Water 1.60 1.60 1.60 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.1 0.9 Resin 2.77 2.77 3.40 0.82 Water 1.49 1.49 1.49 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.2 0.8 Resin 2.58 2.58 3.14 0.82 Water 1.37 1.37 1.37 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.3 0.7 Resin 2.39 2.39 2.88 0.83 Water 1.26 1.26 1.26 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.4 0.6 Resin 2.21 2.21 2.62 0.84 Water 1.14 1.14 1.14 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.5 0.5 Resin 2.02 2.02 2.37 0.85 Water 1.03 1.03 1.03 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.6 0.4 Resin 1.83 1.83 2.11 0.87 Water 0.91 0.91 0.91 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.7 0.3 Resin 1.64 1.64 1.85 0.89 Water 0.80 0.80 0.80 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.8 0.2 Resin 1.45 1.45 1.59 0.91 Water 0.68 0.68 0.68 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.9 0.1 Resin 1.27 1.27 1.34 0.95 Water 0.57 0.57 0.57 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 0.95 0.05 Resin 1.17 1.17 1.21 0.97 Water 0.51 0.51 0.51 1.00(14)

Polyethylene 4.94 5.06 5.06 1.00 1.0 0.0 Resin 1.08 1.08 1.08 1.00 Water 0.45 0.45 0.45 1.00(14) 14 For water, the value is set to 1.0.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.5 Appendix Attachment 4.5-1 EPDM Temperature Specifications

[Ref. 10]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 4.5-2 Seal Material EPDM Working Temperature

[Ref. 11]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 4.5-3 Seal Material EPDM Helium gas permeation rate

[Ref. 11]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 4.5-3 Seal Material EPDM Helium gas permeation rate (Continued)

[Ref. 11]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Attachment 4.5-4 Seal Material EPDM Characteristics With Respect to Damage by Radiation and Hardness Concerns

[Ref. 11]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 .5-5 Additional Support Information about EPDM Resistance to Radiation 8

Up to 5x10 Rads While Retaining Reasonable Flexibility and Strength, Hardness and Very Good Compression Set Resistance

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 .5-5 Additional Support Information about EPDM Resistance to Radiation Up to 5x108 Rads While Retaining Reasonable Flexibility and Strength, Hardness and Very Good Compression Set Resistance (Continued)

[Ref. 12]

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 4.6 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL, dated March 7, 2012 and the following specific Sections:

71.31(a)(1) 71.31(a)(2) 71.33 71.35(a) 71.31(c) 71.43(c) 71.73(d) 71.4(e) 71.71 71.43(f) 71.51(a)(1) 71.43(h) 71.51(a)(2) 71.51 71.63 71.63(b) 71.51(a)(1) 71.51(a)(2) 71.73

3. ANSI N14.5-2014, "American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

4. U.S. Nuclear Regulatory Commission, Containment Analysis for Type B Packages Used to Transport Various Contents, NUREG/CR-6487, November 1996.

5. Particle-size distribution and packing fraction of geometric random packings. H. J. H.

Brouwers, Physical Review E 74, 031309, 2006.

6. DOE Handbook, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, Volume I - Analysis of Experimental Data, U.S. Department of Energy, Washington, D.C., 1994.

7. NUREG/CR-6487, "Containment Analysis for Type B Packages Used to Transport Various Contents," Anderson, B., Carlson, R., & Fischer, L., Lawrence Livermore National Laboratory, Livermore, CA, November 1996, Retrieved on August 28, 2013, Retrieved from http://rampac.energy.gov/docs/nrcinfo/NUREG_6487.pdf.

8. RTL-001-CALC-TH-0102, Rev. 6, "RT-100 Cask Maximum Normal Operating Pressure Calculation" (PROPRIETARY)

9. RTL-001-CALC-TH-0202, Rev. 6, "RT-100 Cask Hypothetical Accident Condition Maximum Pressure Calculation" (PROPRIETARY)

10. Parker O-Ring Handbook ORD 5700, Retrieved on August 28, 2013, Retrieved from http://www.parker.com/literature/ORD%205700%20Parker_O-Ring_Handbook.pdf.

11. TRELLEBORG Sealing Solutions O-Ring and Backup Rings Catalog, August 2011 Edition

12. Glenn Lee, Radiation Resistance of Elastomers, IEEE Transactions on Nuclear Science, Vol.

NS-32, No. 5, October 1985.

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13. Lamarsh, J. & Baratta, A., Introduction to Nuclear Engineering, 3rd Edition, 2001.

14. Fundamentals of Fluid Mechanics, B. Munson, D. Young and T. Okiishi, 5th ed., John Wiley

& Sons, Inc., 2006.

15. Brookhaven National Laboratory, Selected Cryogenic Data Notebook, August 1980.

16. NUREG/CR-6673, "Hydrogen Generation in TRU Waste Transportation Packages,"

Anderson, B., Sheaffer, M., & Fischer, L., Lawrence Livermore National Laboratory, Livermore, CA, May 2000.

17. "Exhibit A of Cask Procurement Agreement dated April 10, 2012 by and between Waste Control Specialists LLC and Robatel Technologies, LLC et al."

18. [Withdrawn]

19. RTL-001-CALC-CN-0101, Rev 6, "Containment Evaluation for the RT-100" (PROPRIETARY)

20. NUREG/CR-4062 Extended Storage of Low-Level Radioactive Waste: Potential Problem Areas, B. Siskind D.R. Dougherty, D. R. MacKenzie.

21. EPRI NP-5977, Radwaste Radiolytic Gas Generation Literature Review, Electric Power Research Institute, September 1988.

22. RT100-REF-01-01, Rev. 0, Historical Cask Summaries by Waste Category (PROPRIETARY)

23. RH-TRU 72-B SAR Payload Appendices, Rev. 0, June 2006.

24. Materials Science and Engineering, W. Callister, Jr., 6th Edition, John Wiley & Sons, Inc.,

2003.

25. 2014-020-CALC-LT-001, Rev. 0, Calculation of RT-100 Pressure Drop Leak Test Conditions (PROPRIETARY)

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5. SHIELDING EVALUATION This Chapter describes the RT-100 shielding evaluation under the RT Quality Assurance Program

[Ref. 1] and summarizes the results to demonstrate compliance with the shielding requirements of 10 CFR 71 [Ref. 2]. The RT-100 cask package is designed to transport radioactive materials including contaminated resin and filter media from nuclear power plant operation, as well as activated hardware (low and high-density hardware). The RT-100 has a robust gamma shielding design comprised of a steel/lead/steel body with a steel primary lid and a steel/lead/steel secondary lid. The primary lid is bolted onto the body, and the secondary lid is bolted into the primary lid.

Both lids, along with their O-ring seals provide secure containment of the radioactive material contents. Analyses presented in this chapter demonstrate that the shielding design produces dose rates below the external radiation requirements of 10 CFR 71 [Ref. 2] under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). The package and vehicle radiation limits are for exclusive use of an open (flat-bed) transport vehicle. Operating limits are established for the specific activities of individual radionuclides (Ci/g) allowed in the contents of the package.

The RT-100 is designed in compliance with the external radiation standards that are specified in 10 CFR Part 71 [Ref. 2] as:

The RT-100 is designed, constructed, and prepared for shipment so that the external radiation levels will not significantly increase under the tests specified in 10 CFR 71.71 (Normal Conditions of Transport) in accordance with 10 CFR 71.43(f) and 10 CFR 71.51(a)(1).

Under NCT tests specified in 10 CFR 71.71, the external radiation levels meet the requirements of 10 CFR 71.47(b) for exclusive-use shipments.

Under HAC tests specified in 10 CFR 71.73, the external radiation level does not exceed 10 mSv/hr (1 rem/hr) at one meter from the surface of the package in accordance with 10 CFR 71.51(a)(2).

The shielding evaluation is based on the descriptions and evaluations presented in the General Information, Structural Evaluation and Thermal Evaluation sections of the application. Results of the shielding evaluation are considered in the preparation of Operating Procedures and the Acceptance Tests and Maintenance Program. An example of information flow for the shielding evaluation is shown in Figure 5-1.

The approach used to calculate the maximum allowable limits is intended to assure that the maximum activity of each radionuclide includes sufficient margin to ensure that the maximum dose rates of a loaded RT-100 cask will comply with regulatory dose rate limits. The process used to calculate the radionuclide-specific source strength densities (Ci/g) is summarized as:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

1. Shielding Model package geometry, materials, source definition, tallies An MCNP6 shielding model is constructed using the minimum shielding material thicknesses, and assumes that the maximum activity concentration of a specific radionuclide completely fills the cask cavity with no credit given for attenuation provided by secondary containers. Separate tallies are defined in the shielding model to determine the peak external dose rate at all locations required by the regulations (i.e. package surface, 1m, 2m). The input file for MCNP6 is a shielding model that represents the package geometry, materials, source definition, and tallies as described in Section 5.3 and 5.4.1.1.

2. External dose rates for package transport conditions The shielding analysis, to determine external dose rates for the RT-100 package, is performed with MCNP6 [Ref. 3]. The output from MCNP6 is a dose rate response in mrem/hr/Ci for each of the generic source energies that has been modified by fluence-to-dose conversion factors and is normalized per 1 Curie of activity. A detailed description of the shielding analysis method is provided in Section 5.4.1.2.

The actual dose rate associated with a particular radionuclide in the cask contents is a function of the package shielding configuration, the interaction of emitted radiation with the cask contents, and the spectrum of radiation emitted from decay of the radionuclides.

Since there are numerous potential radionuclides in the cask contents (as discussed in Section 5.2) with each radionuclide having an emission spectrum containing multiple energy levels, it is impractical to explicitly analyze each energy level for each potential radionuclide separately. Rather, a finite number of energy levels are selected that are representative of the expected range of radiation energies from all radionuclides. The shielding analysis is performed for each of these representative energy levels, referred to in this evaluation as generic energies. The generic energy source term used for the calculation assumes the probability of particle emission is 100 percent per disintegration, and the actual emission probability (i.e. intensity) is accounted for in the dose rate response calculation for individual nuclides.

3. Dose rate response for individual radionuclides The MCNP6 shielding evaluation calculates the dose rate responses for a finite set of generic source energies, independent of the radionuclide responsible for the particle emission. In this evaluation, 26 discrete generic energies were analyzed. The generic energy dose rate responses from MCNP6 are processed and used as input to calculate the dose rate response for individual radionuclides. The method utilized to determine the dose rates from 1 Curie of a specific radionuclide uses the calculated responses of the generic energy lines to determine the response at each energy line of the respective radionuclide.

A radionuclide dose rate response in mrem/hr/Ci is calculated by summing the determined dose rate responses for all energy lines of the respective radionuclide.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 generic energy line response method, with individual MCNP6 inputs modeling the actual source spectrum of each radionuclide. The radionuclides considered individually in this analysis are: Co-60, Zn-65, Fe-59, Mn-54, Co-58, Ag-110m, Cs-134, and Cs-137. A detailed description of the method used to calculate radionuclide dose rate responses is provided in Section 5.4.1.2.

4. Calculate the maximum allowable source strength density Based on the maximum possible mass of resin material in the RT-100, the regulatory dose rate at a given location, and the calculated radionuclide dose rate response at the respective regulatory location, a maximum quantity of each radionuclide allowed to be transported in the cask is specified in the form of specific activities (Ci/g), also referred to as the source strength density. The radionuclide-specific source strength densities are used to verify that the actual contents being shipped will comply with the regulatory requirements for external radiation levels. The maximum allowed source strength densities calculated in the shielding evaluation are used in an operating procedure, referred to as the loading table, which is used for mixtures of radionuclides. A detailed description of the method used to calculate the maximum allowable source strength is provided in Section 5.4.1.3.

The actual contents can vary significantly for shipments; therefore, it is not practical to explicitly evaluate the actual contents that are shipped. This methodology is used to imply a dose rate associated with a particular mixture of radionuclides in the contents by calculating an effective source strength density that is a sum of the fraction of the maximum allowable source strength densities for the individual radionuclides. A sum of the fractions for radionuclide-specific maximum allowable source strength densities that is less than or equal to 1.0 implies that the external radiation levels meet the regulatory limits. Further description of the loading table operating procedure is provided in Chapter 7, Section 7.6.1 of this SAR.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 5-1 Information Flow for the Containment Review Structural General Thermal Evaluation Information Evaluation

Deformation

Dimensions

Combustion

Crushing/Puncture

Materials

Decomposition

Extrusion

Contents

Dehydration

Slump

Exclusive/

Melting

Displacement of Nonexclusive use Contents and Shielding Shielding Review Source Terms Attenuation Dose Rates

Gamma

Material Properties

Gamma

Modeling

Shielding Analysis

Gamma Operating Acceptance Tests Procedures and Maintenance

Closure

Fabrication Requirements Verification

Assembly Leakage Rate Verification

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.1 Description of Shielding Design A description of the shielding design, as well as a summary of the results of the analysis for this design is provided below.

5.1.1 Design Features The RT-100 body is a right circular cylinder 2060 mm in diameter and 2324 mm in height without the impact limiters attached. The cavity of the RT-100 is 1730 mm in diameter and 1956 mm in height. Surrounding the cavity are the cask radial wall, cask bottom wall, primary lid and a secondary lid embedded in the primary lid. The cask radial wall comprises 30 mm of steel, 90 mm of lead and 35 mm of steel for gamma shielding. The cask bottom wall comprises 50 mm of steel, 75 mm of lead and 30 mm of steel for gamma shielding. The primary lid comprises 210 mm of steel for gamma shielding. Embedded in the primary lid is a secondary lid that comprises 100 mm of steel, 60 mm of lead and 10 mm of steel for gamma shielding. The primary lid is 2016 mm in diameter and the secondary lid is 1000 mm in diameter. Under transport conditions, the top and bottom impact limiters provide additional gamma shielding from the 10 mm inner steel shell, polyurethane foam and 4 mm outer steel shell.

Dimensional tolerances and material densities used in the shielding evaluations are given in Section 5.3.1 and Section 5.3.2, respectively.

During normal conditions of transport, shielding evaluations assume that the RT-100 is transported on a truck trailer that is 2438.4 mm wide and 12801.6 mm long with the cask tied down in the center. Thus, the 2 meter radial surface is 3219.2 mm from the cask centerline and the distance to the cab, taking into account the trailer hookup and the distance to back of cab, is 8915.4 mm from the cask centerline. This is a conservative assumption since the actual occupied position is at least 1828.8 mm forward from the back of the cab. A visual representation of the dose locations is provided in Figure 5.4.4-4.

5.1.2 Summary Table of Maximum Radiation Levels The transport regulations provide dose limits in 10 CFR 71.47 and 71.51 at locations external to the package for rates for both Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). A full discussion of the methods employed to analyze the RT-100 cask design and results of applying these methods that demonstrate compliance with the regulatory limits are presented in the sections that follow. Table 5.1.2-1 shown below summarizes the calculated results for the maximum radiation levels allowed for exclusive use shipment using an open (flat-bed) transport vehicle under NCT and HAC for the worst-case loading of radionuclides in spent resin/filter material. These results represent the maximum dose rates for the worst-case allowable contents as presented in Section 5.4.4.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.1.2-1 Resin/Filter Summary Table of External Radiation Levels (Exclusive Use)

Vehicle Normal 2 Meter from 1 Occupied Conditions of Package Surface Edge of Vehicle Position Transport mSv/hr (mrem/hr) mSv/hr mSv/hr (NCT) (mrem/hr)

(mrem/hr)

Radiation2 Side Bottom Top Side Cab Gamma 1.13 (113.2) 0.47 (47.2) 0.87 (86.7) 0.095 (9.5) 0.01 (1.3) 10 CFR 71.47 2.0 (200.0) 2.0 (200.0) 2.0 (200.0) 0.1 (10.0) 0.02 (2.0)

(b) Limit Hypothetical Accident 1 Meter from Package Surface1 Conditions mSv/hr (mrem/hr)

(HAC)

Radiation2 Side Bottom Top Gamma 9.5 (950) 1.16 (115.9) 2.36 (235.7) 10 CFR 71.51 (a)(2) Limit 10 (1000) 10 (1000) 10 (1000)

Note 1: The gamma dose rates are each calculated for a contents limit corresponding to a limiting regulatory dose rate. These values are the maximum of all the radionuclides that were evaluated in Section 5.4.4.5.

Note 2: Typical contents will not contain a significant neutron source term.

5.2 Source Specification The RT-100 is designed to transport nuclear plant radioactive resins/filters and activated hardware.

This content is described in Chapter 1, Section 1.2.2. The radionuclides in these resins, filters, and activated hardware wastes produce primarily gamma emissions and trace neutron emissions from actinide spontaneous fission and alpha-n reactions in the media.

The shielding evaluation for the RT-100 calculates a dose rate response that is normalized to one Curie (mrem/hr/Ci) for each radionuclide with a half-life greater than 1 day. Thus, the gamma dose rate responses are specified on a per Curie basis. The total dose rate will be based on the loaded activity, in Curies, of the resin or filter media. In general, the gamma source terms decrease over time.

5.2.1 Gamma Source Gamma spectra, i.e. the photon lines, are explicitly evaluated in the RT-100 shielding evaluation for radionuclides with greater than 1 day half-life. The radionuclide gamma spectra and intensities are taken from the SCALE 6.0 ORIGEN-S data libraries: origen.rev02.mpdkxgam.dat [Ref. 4].

This data file was reformatted into an Excel file for use in the generic energy dependent response approach to compute radionuclide dose rates from a one Curie source for each radionuclide.

The ORIGEN-S radionuclide data library origen.rev04.endfdec.data [Ref. 5] was read to determine all radionuclides with greater than 1 day half-life for use in the generic approach to compute radionuclide specific gamma dose rates. The list of 281 radionuclides with greater than 1 day half-Robatel Technologies, LLC Page 5-6

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 life is given in Table 5.7.1-1. The selection details and associated spectra data file are provided in the ORIGEN-S: Data Libraries [Ref. 4]. Due to the magnitude of the data, all gamma source terms cannot be provided in the SAR, but the gamma spectra for 1 Curie Co-60 is shown in Table 5.2.1-1. Co-60 is the dominant gamma emitter and dose rate contributor from resins and filter media.

Table 5.2.1-1 One Curie Co-60 Gamma Source Term Photon Energy¹ Intensity 1 Curie (MeV) Photon/dis Photon/s MeV/s 1.17E+00 1.00E+00 3.70E+10 4.34E+10 1.33E+00 1.00E+00 3.70E+10 4.93E+10 Total 7.40E+10 9.27E+10 Note 1: Only the prominent photon lines are considered for Co-60 in the individual analysis.

Beta Emitter/ Bremsstrahlung source Another source of gamma radiation is from radionuclide beta emission Bremsstrahlung (braking radiation). Contributions from Bremsstrahlung gamma radiation have been evaluated by explicit mode e-p (electron-photon) transport calculation. The following radionuclides with beta Emax > 2 MeV were evaluated, as well as Cs-137, due to its typical high activity in resins and filters:

Y-90 Emax = 2.281 MeV Sb-124 Emax = 2.302 MeV Cs-137 Emax = 1.175 MeV La-140 Emax = 2.165 MeV Ce-144 Emax = 2.996 MeV Binned radionuclide beta source spectra are compiled in Calculation Package RTL-001-CALC-SH-0101, Rev. 1 [Ref. 6]. As discussed in Section 5.4.4.3, an assessment of the contribution to exterior dose rates from Bremsstrahlung due to these fission product beta emitting radionuclides is evaluated for the RT-100 NCT configuration and found to contribute less than 1.0 percent of the total dose due to gamma radiation. Therefore, gamma radiation from Bremsstrahlung is not included in the determination of maximum allowed source strength densities.

5.2.2 Neutron Source The RT-100 cask is not designed for shielding neutrons, thus neutron emitters in the contents are limited to trace amounts that may be present in the activated resin and filter media. For this packaging, any neutron source is limited to 3.5E-06 Ci/g, based on Class C burial limits.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.3 Shielding Model MCNP6 [Ref. 3] is used to perform the shielding evaluation of the RT-100. Two sets of MCNP6 shielding models are created for the evaluations of the RT-100 for NCT and HAC. In both cases, the model geometry was developed from the drawings provided in Appendix 1.4. For evaluation purposes, the thicknesses of shielding materials that comprise the package (steel shells, lead shielding, and lids) are reduced by subtracting the manufacturing tolerance from the nominal dimensions. Using minimum thicknesses of shielding materials in the model bounds any effect that a variation in thickness due to fabrication tolerances may have on external dose rates. A summary of the nominal and minimum shield thicknesses is given in Table 5.3-1. For the NCT Proprietary Information Content Withheld Under 10 CFR 2.390(b) responses. The shielding from the high integrity container (HIC) used to store and transfer resin into the RT-100 cavity is neglected in the shielding evaluations. The effects of resin and filter density changes and redistribution of the content media due to NCT and HAC are modeled by decreasing the volume occupied by the source term.

Table 5.3-1 Model Shielding Thicknesses Nominal Model Component (mm) (mm)

Interior Barrel 30 29.7 Lead 90 85 Exterior Barrel 35 34.7 Bottom Forging 50 49.7 Bottom Lead 75 70 Bottom Wall 30 29.7 Primary Lid 210 209.5 Secondary Lid 170 169.5 Secondary Lid Lead 60 58 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.3.2 Material Properties Materials used for the fabrication of the RT-100 are stainless steel and lead in the body, and for the fabrication of the impact limiters the materials used are polyurethane and stainless steel. The material properties used in the shielding evaluations are shown in Table 5.3.2-1. No changes in material properties are expected under Normal Conditions of Transport or Hypothetical Accident Conditions. Melting of the lead will not occur based on the thermal evaluation in Chapter 3. Also, the shielding properties of these materials will not degrade during the service life of the RT-100.

Contents transported in the RT-100 are resins and filter media. The following four materials, typical of resins and filter media, are considered as the cask contents:

o Polystyrene based resins such as Duralite o Activated Charcoal o Nylon filter media o Zeolite - hydrated aluminosilicates such as Faujasite In the case of nylon filter media, any steel cartridge structure is neglected and the cavity is assumed to be completely filled with nylon material at a reduced density. It is established in CN-13039-502

[Ref. 8] that carbon (activated charcoal) is the least effective shielding material between the four materials considered, and thus calculates the most restrictive specific activity (Ci/g) limits for all radionuclides, so in the shielding analysis the contents are modeled as carbon for both NCT and HAC.

The densities for resin and filter materials can vary quite a lot depending on the size of the particles, configuration of the filter media (i.e. tubes or fiber sheets) and the theoretical density of the material. The theoretical densities of polystyrene, carbon, Faujasite and nylon are 1.04, 2.2, 1.93 and 1.08 g/cm3, respectively, but resin media bulk densities for all the typical material compositions can vary from 0.3 to 1.0 g/cm3. It is established in CN-13039-502 [Ref. 8] that a contents density of 1 g/cm³ calculates the most restrictive specific activity (Ci/g) limits for all radionuclides, so the shielding analysis uses this density for the modeled contents media for both NCT and HAC.

For all bottom dose rate response calculations, a 10% compaction of the resin is considered, and the results from the shielding analysis are adjusted to consider this variation, though the bottom dose rate locations are never the limiting case.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.3.2-1 RT-100 Material Composition Summary Density Nuclide Weight Material (g/cm3) Element ID Fraction Stainless Steel 304 7.94 Fe 26000 0.68375

[Ref. 9] Ni 28000 0.09500 Cr 24000 0.19000 Mn 25055 0.02000 Si 14000 0.01000 C 6000 0.00080 P 15000 0.00045 Lead [Ref. 9] 11.35 Pb 82000 1.0000 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

Polystyrene 0.3 - 1.0 H 1001 0.0774

[Ref. 10] C 6000 0.9226 Activated Carbon 0.3 - 1.0 C 6000 1.0000

[Ref. 9]

Nylon 0.3 - 1.0 H 1001 0.0980

[Ref. 10] C 6000 0.6369 N 7014 0.1238 O 8016 0.1414 Zeolite 0.3 - 1.0 O 8016 0.6067 (Faujasite-Na) Si 14000 0.2263

[Ref. 11] Al 13027 0.0895 H 1001 0.0306 Na 11023 0.0229 Ca 20000 0.0199 Mg 12000 0.0040 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4 Shielding Evaluation - Resins and Filters Section 5.4 describes the shielding evaluation for the RT-100 containing resin and filter contents using industry accepted methods. The shielding evaluations for mass restricted filters and activated hardware are outlined in Sections 5.5 and 5.6, respectively.

Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b) 5.4.2 Input and Output Data All relevant inputs and outputs for the gamma shielding analysis are provided with calculation package CN-13039-502. Post processing of the energy dependent responses into detailed dose rate responses (mrem/hr/Ci) is performed for all radionuclides and is shown in Tables 5.7.2-1 and 5.7.2-2. Using these responses and the content activity loading, the total dose rate in mrem/hr can be computed for NCT and HAC conditions. This is demonstrated in Section 5.4.4.5 for the maximum source strength density of Co-60. A guide is provided in Appendix 5.7.4 that relates the process steps to the input and output files used for the shielding evaluation.

5.4.3 Flux-to-Dose Rate Conversion MCNP6 [Ref. 3] calculates a photon flux (particles/s-cm2) at a particular tally or detector location given the source magnitude. These values are converted into dose by use of flux-to-dose response functions. This conversion is done internally in MCNP6 [Ref. 3] by associating dose response functions to each tally in the input file. Gamma flux-to-dose response functions used in these calculations are listed in Table 5.4.3-1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.3-1 ANSI/ANS 6.1.1-1977 - Gamma Flux-to-Dose Conversion Factors [Ref. 13]

(rem/hr)/

Gamma Energy (MeV)

(photon/cm2-s) 0.01 3.96E-06 0.03 5.82E-07 0.05 2.90E-07 0.07 2.58E-07 0.10 2.83E-07 0.15 3.79E-07 0.20 5.01E-07 0.25 6.31E-07 0.30 7.59E-07 0.35 8.78E-07 0.40 9.85E-07 0.45 1.08E-06 0.50 1.17E-06 0.55 1.27E-06 0.60 1.36E-06 0.65 1.44E-06 0.70 1.52E-06 0.80 1.68E-06 1.00 1.98E-06 1.40 2.51E-06 1.80 2.99E-06 2.20 3.42E-06 2.60 3.82E-06 2.80 4.01E-06 3.25 4.41E-06 3.75 4.83E-06 4.25 5.23E-06 4.75 5.60E-06 5.00 5.80E-06 5.25 6.01E-06 5.75 6.37E-06 6.25 6.74E-06 6.75 7.11E-06 7.50 7.66E-06 9.00 8.77E-06 11.0 1.03E-05 13.0 1.18E-05 15.0 1.33E-05 Robatel Technologies, LLC Page 5-25

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4.4 External Radiation Levels The maximum external radiation levels are determined by the quantity of each radionuclide in the resin and filter media that is to be shipped. This limiting quantity of each radionuclide is determined by the respective source strength density limit of each. For the radionuclides considered, the source strength density is always limited by either the NCT 2 meter or the HAC side 1 meter regulatory limits. Thus, the maximum dose rate that can be measured at any regulatory location can only be equal to the regulatory limit at the NCT 2 meter or the HAC side 1 meter locations. For example, for radionuclides whose source strength density is limited by the NCT 2 meter location, the maximum dose rate at 2 meters from the edge of the vehicle is 9.5 mrem/hr and the dose rate at all other regulatory locations will be some amount less than the regulatory limit for that location because the source strength density is based on the NCT 2 meter location.

Likewise, for radionuclides whose source strength density is limited by the HAC side 1 meter location, the maximum dose rate at 1 meter from the package surface is 950 mrem/hr and the dose rate at all other regulatory locations will be some amount less than the regulatory limit at that location.

Table 5.7.2-1 and Table 5.7.2-2 give the complete list of gamma radionuclide responses. Using these responses and the respective source strength density limits in Table 5.7.3-1, the total dose rate in mrem/hr at each regulatory location can be computed for any radionuclide as described in Section 5.4.4.5. The maximum dose rates under NCT and HAC from each radionuclide individually are shown in Table 5.4.4-5 and Table 5.4.4-6. Table 5.4.4-1 provides a summary of the maximum calculated dose rate at each regulatory location, and the radionuclide responsible for each maximum dose rate. Combining contributions from multiple radionuclides, as is done in the loading table, can only result in a dose rate at each regulatory location that is equal to or less than the maximum values reported below in Table 5.4.4-1.

Table 5.4.4-1 Maximum Dose Rates and Responsible Radionuclides 2 Meters from Vehicle Normal Edge of Occupied Conditions Package Surface (mrem/hr)

Vehicle Position of Transport (mrem/hr) (mrem/hr)

Source Side Bottom Top Side Cab 113.2 47.2 86.7 9.50 1.21 Gamma Pm-144 Cs-144 Pm-144 MULTIPLE MULTIPLE Hypothetical Accident 1 Meter from Package Surface (mrem/hr)

Conditions Source Side Bottom Top 950.0 115.93 235.7 Gamma MULTIPLE Mn-54 Mn-54 Robatel Technologies, LLC Page 5-26

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4.4.1 MCNP6 Statistics Evaluation 5.4.4.1.1 Tally Statistics Diagnostics MCNP6 produces information about a simulation to allow the user to assess the precision (not the accuracy) of the result. MCNP6 provides 10 statistical tests performed on the tally for assessing the reliability of results. If any of the 10 tests are not met, they are recorded as NO. MCNP6 automatically produces additional output to aid the user in interpreting the cause of the failed tests.

The 10 tests are summarized as, Tally Mean, x

1. The mean must exhibit, for the last half of the problem, only random fluctuations as N increases. No up or down trends must be exhibited.

Relative Error, R

2. R must be less than 0.1 (0.05 for point/ring detectors).

3. R must decrease monotonically with N for the last half of the problem.

4. R must decrease as 1/N for the last half of the problem.

Variance of the Variance, VOV:

5. The magnitude of the VOV must be less than 0.1 for all types of tallies.

6. VOV must decrease monotonically for the last half of the problem.

7. VOV must decrease as 1/N for the last half of the problem.

Figure of Merit, FOM:

8. FOM must remain statistically constant for the last half of the problem.

9. FOM must exhibit no monotonic up or down trends in the last half of the problem.

Tally PDF, f(x)

10. The SLOPE determined from the 201 largest scoring events must be greater than 3.

The MCNP Manual states that A tally is considered to be converged with high confidence only when it passes all ten statistical checks [Ref. 3]. Every tally in all MCNP6 outputs in the shielding analysis passes all 10 statistical tests, indicating that the reported results are reliable.

5.4.4.1.2 Fractional Standard Deviation of Individual Tally Segments Of all tally bins included in the 72 MCNP outputs for the gamma shielding analysis, there are a total of 17 bins with reported fractional standard deviations larger than the MCNP6 requirement for reliable results (fsd<0.10). These tally segment bins with a fractional standard deviation greater than 0.1 are far from the maximum segment, with a calculated dose rate that is orders of magnitude less than the value reported at the maximum segment. This is shown for Cs-137 in Figure 5.4.1-1, where it can be noted that the tally segments that dont meet the fsd requirement are on the bottom impact limiter, with reported dose rates that are multiple orders of magnitude less than the segments at the axial location of the streaming peak. All maximum dose rate segments have an fsd of 0.05 or less, most being significantly lower.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4.4.2 Media Composition and Density The effect of the media composition is addressed in calculation RTL-001-CALC-SH-0201 Section 7.7.2 Content Density and Material Variation [Ref. 7]. The MCNP calculations for the final dose rate response functions of the 8 radionuclides considered individually are performed using carbon at a density of 1.00 g/cm3 as the material composition and density. Parametric studies were performed to evaluate the effect of other media compositions (polystyrene, nylon, and zeolite) and media densities in the range from 0.65 to 1.0 g/cm3 on the dose rate response. The maximum allowable source strength density (Ci/g) decreases slightly with increasing material density and carbon material composition results in the most limiting source strength density. The final dose rate responses used to calculate the maximum source strength densities are calculated modeling the resin media as carbon at 1 g/cm³, instead of adjusting the calculated dose rate responses for polystyrene at 0.65 g/cm³ using correction factors as originally discussed in Section 7.7.2 of the calculation RTL-001-CALC-SH-0201 [Ref. 7].

5.4.4.2.1 Effect of Media Composition The effect of media composition was evaluated by calculating the dose rate response for the generic energies using the typical resin and filter material compositions at a fixed material density, 0.65 g/cc. The ratio X(r,) between the dose rate calculated for each media composition and the base case (polystyrene) is determined for each radionuclide X at each regulatory dose rate limit location r. A comparison of the average of the calculated ratios is presented in Table 5.4.4-2, where it is shown that modeling activated carbon as the attenuating media results in the highest calculated dose rates.

(, ) [ ]

(, ) =

(, = ) [ ]

() = { (, )}

Table 5.4.4-2 Media Composition Comparison

()

Side Bottom Top Polystyrene 1.00 1.00 1.00 Activated Carbon 1.07 1.10 1.09 Nylon 0.98 1.02 1.05 Zeolite [Ref. 10] 1.04 1.03 1.06 Robatel Technologies, LLC Page 5-29

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4.4.2.2 Effect of Media Density The effect of media density was evaluated by calculating the specific activity limit aX for the generic energies using the material compositions for a range of material densities for one material composition, polystyrene. The ratio X(r,) between the specific activity limit calculated for each density and the base case (0.65 g/cm³) is determined for each radionuclide X at each regulatory dose rate limit location r. A comparison of the maximum calculated ratios is presented in Table 5.4.4-3, where it shown that modeling the content media at 1 g/cm³ results in the most restrictive calculated specific activity limits.

(, )[]

(, ) =

(, = 1.00)[]

() = { (, )}

Table 5.4.4-3 Media Density Comparison

[g/cm3] ()

0.65 1.053 0.75 1.041 0.85 1.035 1.00 1.000 Robatel Technologies, LLC Page 5-30

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.4.4.3 Shielding Evaluation Uncertainty Accurate results from the shielding analysis of transport casks are important to ensure that loading limits yield doses that do not exceed the regulatory limits for external radiation levels. Content loading limits (i.e. specific source strength densities) for the RT-100 are determined using regulatory dose rate limits. These regulatory limits are reduced by 5 percent to account for any uncertainty in representing the packaging and the characterization of the contents. The uncertainties and margins are summarized in Figure 5.4.4-2.

Regulatory limit for external radiation levels (71.47 and 71.51) 5.0% Regulatory limits are reduced to account for any uncertainty in representing the packaging and the characterization of the contents Radiation source generation MCNP Tally+2 1.0% Bremsstrahlung 1.5% Rounding up actual decay energies (Section 5.4.4.3.1.1) (Section 5.4.4.3.1.1)

Calculated dose rate Attenuation from other material (i.e. secondary containers) not included 5.0%

in shielding analysis (Section 5.4.4.3.2) 1.5% Nominal media bulk density (Section 5.4.4.3.3)

Measured external radiation levels Figure 5.4.4-2 Summary of Calculated Dose Rate Margins 5.4.4.3.1 Calculation of Dose Rates Uncertainty in calculated dose rates arises from three major analysis areas: radiation source generation, use of cross-section data, and the radiation transport codes used to evaluate doses.

5.4.4.3.1.1 Radiation Source Generation Generation of the source specification is discussed in detail in Section 5.2.

Bremsstrahlung The dose rate from Bremsstrahlung is not included, as this source term accounts for less than 1%

of the regulatory dose rate limit, and there is no significant uncertainty associated with the radiation source definition for gammas.

An evaluation of the contribution of Beta/Bremsstrahlung radiation to external dose rates was performed in the calculation report RTL-001-CALC-SH-0201. The NCT results for this analysis Robatel Technologies, LLC Page 5-31

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 are presented for Y-90, Sb-124, Cs-137, La-140 and Ce-144 in Table 5.4.4-4. Based on this assessment, the contribution of Beta/Bremsstrahlung to exterior doses on the RT-100 can be neglected as long as the beta activity is well below 60,000 Curies (the Curie content of Ce-144 that would produce 1% of the chosen dose rate limit for two meters, or 0.095 mrem/hr). The 3000 A2 limit precludes the shipment of sufficient Curies of the beta/bremsstrahlung emitting isotopes to exceed the 1% of the total dose rate limit at two meters, and the bremsstrahlung gamma response functions are therefore omitted from the loading tables.

Table 5.4.4-4 NCT Dose Rate Responses Due to Bremsstrahlung Maximum Response + 2 (mrem/hr/Ci) for Side Dose Locations Radionuclide Top Surface Side Surface Bottom Surface 2 Meter Cab Y-90 2.80E-06 6.00E-06 3.78E-06 6.60E-07 1.88E-07 SB-124 2.00E-06 1.43E-06 4.49E-07 1.59E-07 4.38E-08 CS-137 0.00E+00 3.37E-11 0.00E+00 1.45E-11 5.89E-12 LA-140 6.13E-08 4.46E-07 3.94E-07 3.88E-08 1.25E-08 CE-144 4.76E-06 1.23E-05 1.80E-05 1.95E-06 5.49E-07 CE-144-void 1.16E-04 2.75E-04 2.43E-04 3.64E-05 1.04E-05 CE-144-4 1.16E-05 2.37E-05 1.25E-05 3.12E-06 8.18E-07 Rounding up actual decay energies The use of a generic energy spectrum method as described in Section 5.4.1 overestimates the dose rate from specific radionuclides. Rounding up the gamma energies of the specific radionuclides to the nearest generic energy line may result in a 1.5% to 40% increase in dose rates depending on the radionuclide.

5.4.4.3.1.2 Cross Section Data The use of continuous pointwise energy cross section data libraries in MCNP6 eliminates any uncertainty due to the choice of energy-collapsing spectrum and parent fine-group data.

5.4.4.3.1.3 Radiation Transport Codes The use of MCNP6 to calculate the dose rate responses is discussed in detail in Section 5.4.1.2.

The mean tally value plus two standard deviations as calculated by MCNP6 is used to calculate the limiting dose rate responses. Adding two standard deviations to the calculated average value means that there is at least a 95% confidence that the calculated dose rate response will be less than the limiting dose rate response.

5.4.4.3.2 Attenuation from other material (i.e. secondary containers) not included in shielding analysis The secondary container is fabricated using approximately 1.27 cm thickness polyethylene. The half-value layer (HVL) for polyethylene for 1 MeV for gammas is approximately 15 cm. The Robatel Technologies, LLC Page 5-32

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 secondary container provides a reduction in external dose levels of approximately 6%.

5.4.4.3.3 Nominal Media Bulk Density The bulk density of dewatered resins is typically less than 1 g/cm3. Compliance with the regulatory dose rate limits is demonstrated using the limiting source strength density values for a bulk density of 1 g/cm3. When the maximum allowed source strength density (i.e. the value at 1 g/cm3) is used to estimate the expected dose rate for media densities less than 1 g/cm3, the calculated dose rate is greater than the estimated dose rate using the source strength density (Ci/g) for 1 g/cm 3, and the actual media density (g/cm3) and associated dose rate response (mrem/hr/Ci). An example of this conservatism is shown in Figure 5.4.4-3.

Figure 5.4.4-3 Example of Media Density Effect 5.4.4.4 Loading Table Ultimately, a loading table specifying the maximum Curies per gram of each radionuclide is developed. The maximum Curies per gram are defined as the minimum specific activity at which the regulatory dose rate limit for either NCT or HAC is met. For conservatism 5% lower than the regulatory limits is assumed, thus the NCT limit at the package surface is taken as 190 mrem/hr, at 2 meters from the transport vehicle as 9.5 mrem/hr and the cab limit is taken as 1.9 mrem/hr.

Similarly, the HAC limit at 1 meter is evaluated at 950 mrem/hr.

For NCT, the maximum source strength density (Ci/g) is computed as the minimum of the following:

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 190 [ ]

( ) [ ] =

( ) [ ] (1.00 []

3 )

9.5 [ ]

(2 ) [ ] =

(2 ) [ ] (1.00 []

3 )

1.9 [ ]

() [ ] =

() [ ] (1.00 []

3 )

For HAC, the maximum source strength density (Ci/g) is computed as the minimum of the following:

950 [ ]

(1 ) [ ] =

(1 ) [ ] (1.00 []

3 )

Whichever condition is more limiting, either NCT or HAC determines the maximum Curies per gram of a particular radionuclide X. Section 5.5, Appendix 5.5.3 gives the complete list of the maximum allowable source strength density (Ci/g) limits for each of the gamma emitting radionuclides and for each of the predominant neutron emitting radionuclides. These limits are based on resin or filter material at a density of 1.0 g/cm3 filling the entire cavity of the RT-100 (4597809 total grams).

In the loading table, the actual source strength density for each radionuclide is divided by the maximum allowable source strength density from the shielding evaluations, and the results summed for all radionuclides must be less than 1.0 to be acceptable for shipment.

5.4.4.5 Dose Rates for Maximum Radionuclide Loading External radiation levels at a given dose rate location X(r) are computed for the maximum radionuclide loadings in the RT-100 by using the calculated nuclide-specific dose rate per Curie responses DX(r) along with the source strength density limit set for the respective radionuclide aX to demonstrate the dose rates produced for a given content isotopic loading as follows:

() [ ] = [ ] () [ ] []

An example of the maximum does rates and their corresponding measurement locations are shown Robatel Technologies, LLC Page 5-34

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 in Figure 5.4.4-4 for Co-60 (assuming the maximum mass of 4597809 grams).

Dose rates for other radionuclides with a greater than one (1) day half-life can also be computed in the same manner using radionuclide gamma dose rate per Ci responses in Table 5.7.2-1 for NCT and Table 5.7.2-2 for HAC, and the source strength densities in Table 5.7.3-1. The estimated dose rates due to the maximum allowed loading of each radionuclide considered individually is shown in Table 5.4.4-5 and Table 5.4.4-6. Any mixture of radionuclides can be evaluated with a characterization of the actual source strength density (Ci/g) of each radionuclide, using the loading table.

The RT-100 loading table procedure as outlined in Chapter 7, Section 7.6.1 provides the guidance on the acceptability of a mixture of radionuclides in resin and filter material based on the maximum Ci/g for each radionuclide in the mixture. This loading table procedure, based on sum of the fractions of contributions, determines whether the radionuclide mixture meets or exceeds 10 CFR 71.47(b) or 10 CFR 71.51(a) (2) [both Ref. 2] radiation limits. In addition, the loading table determines whether the combined A2 and mixture heat load meets or exceeds licensed limits.

Top - 30.50 mrem/hr Cab - 1.16 mrem/hr Side - 64.43 mrem/hr 8.92 meters Bottom - 38.84 mrem/hr Vertical Plane of Transport Vehicle 9.50 mrem/hr 2 meters Figure 5.4.4-4 NCT Maximum Gamma Dose Rates for Co-60 Content at 1.00 g/cc Robatel Technologies, LLC Page 5-35

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab na24 3.38E-06 65.9 46.7 33.6 9.5 1.2 bi208 3.76E-06 65.9 47.1 34.2 9.5 1.2 cs144 5.27E-06 65.2 47.2 35.1 9.5 1.2 y88 1.02E-05 67.0 44.3 29.1 9.5 1.2 la140 1.63E-05 66.5 43.5 29.6 9.5 1.2 sb124 2.19E-05 66.9 43.5 29.9 9.5 1.2 eu156 2.26E-05 65.8 45.1 30.9 9.5 1.2 sc48 2.67E-05 66.2 43.4 30.8 9.5 1.2 la138 3.87E-05 67.0 42.7 29.6 9.5 1.2 tb156 4.10E-05 67.1 43.7 30.4 9.5 1.2 ag106m 4.45E-05 67.2 43.3 30.9 9.5 1.2 lu169 4.55E-05 66.5 43.6 30.1 9.5 1.2 na22 5.38E-05 68.3 43.9 31.7 9.5 1.2 sb120m 5.51E-05 66.6 43.9 33.4 9.5 1.2 i124 5.54E-05 66.9 44.0 30.5 9.5 1.2 br82 5.93E-05 67.4 43.0 31.5 9.5 1.2 lu172 6.68E-05 66.4 44.2 32.9 9.5 1.2 ta182 6.93E-05 67.3 43.6 31.5 9.5 1.2 ca47 7.23E-05 67.8 44.0 31.5 9.5 1.2 sc46 7.66E-05 66.3 43.6 33.9 9.5 1.2 te131m 7.96E-05 66.7 44.6 32.4 9.5 1.2 eu152 8.35E-05 67.0 43.1 31.0 9.5 1.2 as72 8.49E-05 66.1 46.0 35.0 9.5 1.2 tm172 9.32E-05 66.7 42.9 29.1 9.5 1.2 eu154 9.69E-05 67.5 44.0 32.5 9.5 1.2 cs136 1.09E-04 66.5 44.9 36.3 9.5 1.2 pm148 1.10E-04 66.9 42.8 29.9 9.5 1.2 ge69 1.25E-04 66.8 43.5 31.4 9.5 1.2 tb160 1.56E-04 66.4 44.0 34.3 9.5 1.2 sn125 1.82E-04 65.7 45.5 32.3 9.5 1.2 rh102m 1.88E-04 69.2 43.7 36.8 9.5 1.2 gd147 1.93E-04 67.4 43.8 33.9 9.5 1.2 tc96 2.23E-04 66.4 45.2 46.3 9.5 1.2 sr83 2.25E-04 67.4 43.8 31.9 9.5 1.2 k40 2.42E-04 66.9 42.7 29.3 9.5 1.2 co60 2.50E-04 64.4 38.8 30.5 9.5 1.2 as76 2.84E-04 67.2 44.6 31.5 9.5 1.2 nb92 3.41E-04 66.3 45.6 42.5 9.5 1.2 np238 3.49E-04 65.7 45.6 38.4 9.5 1.2 am240 4.09E-04 64.7 46.1 43.0 9.5 1.2 tb158 4.32E-04 65.4 45.5 40.2 9.5 1.2 pm148m 4.79E-04 74.0 43.8 46.2 9.5 1.2 ag110m 4.84E-04 64.5 38.7 30.6 9.5 1.2 ho166m 5.45E-04 69.8 44.7 49.8 9.5 1.2 pa232 5.96E-04 65.0 46.0 44.8 9.5 1.2 sb126 6.14E-04 83.0 42.1 56.4 9.5 1.2 Robatel Technologies, LLC Page 5-36

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab nb94 6.40E-04 68.4 46.6 60.1 9.5 1.2 rb84 6.41E-04 66.0 45.2 42.9 9.5 1.2 tm168 6.65E-04 67.8 45.2 49.0 9.5 1.2 pr144 6.75E-04 65.7 45.5 31.1 9.5 1.2 fe59 7.11E-04 64.5 39.0 31.7 9.5 1.2 rh99 7.83E-04 69.1 42.7 32.1 9.5 1.2 tm165 8.21E-04 68.0 43.0 34.4 9.5 1.2 zr89 1.18E-03 69.6 42.5 30.4 9.5 1.2 rh102 1.23E-03 69.8 42.5 33.5 9.5 1.2 as71 1.37E-03 72.7 42.5 37.6 9.5 1.2 ru106 1.57E-03 69.1 43.9 34.2 9.5 1.2 tc95m 1.59E-03 69.2 44.8 51.4 9.5 1.2 rb86 1.82E-03 66.3 45.4 36.7 9.5 1.2 tc98 1.97E-03 99.9 42.0 86.1 9.5 1.2 ag108m 2.12E-03 102.2 40.9 86.4 9.5 1.2 ho166 2.17E-03 66.5 42.8 28.5 9.5 1.2 ir194m 2.25E-03 109.2 34.4 71.0 9.5 1.2 zn65 2.34E-03 64.3 39.2 35.1 9.5 1.2 cs132 2.34E-03 87.1 39.7 51.4 9.5 1.2 lu171 2.56E-03 80.4 44.7 65.0 9.5 1.2 nb95 2.59E-03 87.3 45.3 79.1 9.5 1.3 zr95 2.62E-03 87.3 45.3 79.1 9.5 1.3 tb153 2.83E-03 70.3 43.5 41.2 9.5 1.2 te121m 3.15E-03 71.1 41.9 34.3 9.5 1.2 ag105 3.32E-03 76.6 42.6 45.3 9.5 1.2 sb127 3.61E-03 89.3 40.5 60.3 9.5 1.2 sb122 3.62E-03 80.2 40.0 41.3 9.5 1.2 pm144 3.14E-03 113.2 25.8 86.7 7.8 1.0 cd115m 4.17E-03 67.0 44.4 33.7 9.5 1.2 kr79 4.27E-03 74.5 41.2 38.7 9.5 1.2 pm146 4.35E-03 92.1 43.4 80.2 9.5 1.3 gd149 4.88E-03 83.2 41.9 59.0 9.5 1.2 i126 5.10E-03 88.0 39.3 53.7 9.5 1.2 cs134 6.03E-03 62.4 37.2 44.7 9.5 1.2 os185 6.57E-03 64.9 46.8 56.3 9.5 1.2 cu64 6.61E-03 77.3 39.5 37.1 9.5 1.2 pm143 6.71E-03 87.3 45.3 79.1 9.5 1.3 co58 6.96E-03 64.3 37.8 43.3 9.5 1.2 as74 7.27E-03 106.9 32.1 61.5 9.5 1.1 ba131 7.28E-03 82.8 39.8 46.8 9.5 1.2 br77 9.13E-03 86.3 39.3 55.7 9.5 1.2 ce143 9.23E-03 82.9 40.8 53.2 9.5 1.2 pm151 1.27E-02 95.0 38.8 70.3 9.5 1.2 mn54 1.65E-02 79.8 38.6 73.6 9.5 1.3 y91 1.80E-02 67.8 44.0 31.3 9.5 1.2 sb125 1.29E-02 96.2 17.8 69.8 6.2 0.8 Robatel Technologies, LLC Page 5-37

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab sr85 1.15E-02 85.0 4.8 54.9 4.5 0.5 rb83 1.23E-02 86.8 6.6 57.0 4.8 0.6 te121 1.15E-02 84.6 4.7 54.6 4.5 0.5 sn123 2.46E-02 66.3 45.4 36.7 9.5 1.2 i131 1.71E-02 72.2 10.5 46.8 4.6 0.5 ag110 3.68E-02 81.2 41.4 46.1 9.5 1.2 cm241 1.16E-02 55.4 1.3 29.6 2.8 0.3 lu176 1.12E-02 52.6 0.2 26.9 2.6 0.3 au198 2.01E-02 66.5 10.1 35.3 4.6 0.5 se75 1.14E-02 52.7 0.3 27.0 2.6 0.3 y87 1.16E-02 52.8 0.3 27.0 2.6 0.3 rh101 1.30E-02 52.6 0.2 26.9 2.6 0.3 in114m 4.87E-02 106.1 37.7 84.3 9.5 1.2 hf181 1.36E-02 53.0 0.4 27.3 2.6 0.3 ru103 1.83E-02 63.4 4.3 37.3 3.5 0.4 cs129 2.97E-02 71.4 13.0 40.2 5.2 0.6 xe127 1.73E-02 52.7 0.3 27.0 2.6 0.3 bi210m 2.20E-02 61.0 3.5 35.1 3.3 0.4 ru97 2.01E-02 54.7 1.5 29.0 2.9 0.3 yb169 1.97E-02 52.6 0.3 26.9 2.6 0.3 co57 1.99E-02 52.9 0.4 27.2 2.6 0.3 ta183 2.04E-02 52.6 0.2 26.9 2.6 0.3 np239 2.04E-02 52.6 0.2 26.9 2.6 0.3 ba133 2.04E-02 52.6 0.2 26.9 2.6 0.3 zr88 2.05E-02 52.6 0.2 26.9 2.6 0.3 zn72 2.06E-02 52.6 0.2 26.9 2.6 0.3 rh101m 2.23E-02 55.4 0.6 29.3 2.8 0.3 cd115 3.40E-02 79.4 4.0 50.0 4.2 0.5 pt191 3.21E-02 71.7 4.6 43.7 3.9 0.4 te132 2.17E-02 52.6 0.2 26.9 2.6 0.3 cf249 2.18E-02 52.7 0.3 27.0 2.6 0.3 ba140 3.43E-02 76.6 3.5 47.6 4.0 0.4 hf182 2.22E-02 52.6 0.2 26.9 2.6 0.3 tc99m 2.24E-02 52.6 0.2 26.9 2.6 0.3 sn117m 2.25E-02 52.6 0.2 26.9 2.6 0.3 te129m 8.36E-02 94.3 40.6 70.6 9.5 1.2 hf175 2.30E-02 52.6 0.2 26.9 2.6 0.3 u235 2.32E-02 52.6 0.3 26.9 2.6 0.3 te123m 2.37E-02 52.6 0.2 26.9 2.6 0.3 hg203 2.45E-02 52.6 0.2 26.9 2.6 0.3 cm247 2.45E-02 52.6 0.2 26.9 2.6 0.3 cf251 2.48E-02 52.6 0.2 26.9 2.6 0.3 ce139 2.50E-02 52.6 0.2 26.9 2.6 0.3 pu246 2.54E-02 52.6 0.2 26.9 2.6 0.3 mo99 9.03E-03 100.0 38.4 76.0 9.5 1.2 ar37 4.49E-02 70.4 7.1 43.9 4.1 0.5 Robatel Technologies, LLC Page 5-38

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab sc47 2.93E-02 52.6 0.2 26.9 2.6 0.3 in113m 3.11E-02 52.6 0.2 26.9 2.6 0.3 u237 3.13E-02 52.6 0.2 26.9 2.6 0.3 cm243 3.16E-02 52.6 0.2 26.9 2.6 0.3 pa233 3.22E-02 52.6 0.2 26.9 2.6 0.3 ga67 4.31E-02 57.0 4.3 31.2 3.4 0.4 tb155 3.46E-02 53.0 0.3 27.3 2.6 0.3 er172 3.72E-02 53.1 0.4 27.4 2.6 0.3 nd147 6.37E-02 82.2 5.6 52.9 4.5 0.5 lu177m 3.73E-02 52.6 0.2 26.9 2.6 0.3 cm245 3.76E-02 52.6 0.2 26.9 2.6 0.3 np236 3.88E-02 52.6 0.2 26.9 2.6 0.3 cs137 7.95E-02 94.2 9.4 65.9 5.3 0.6 cu67 4.00E-02 52.6 0.2 26.9 2.6 0.3 au199 4.09E-02 52.6 0.2 26.9 2.6 0.3 ce141 4.14E-02 52.6 0.2 26.9 2.6 0.3 tm167 4.47E-02 54.7 0.5 28.7 2.7 0.3 th227 5.64E-02 53.3 0.8 27.5 2.7 0.3 ra223 5.77E-02 53.4 0.5 27.6 2.6 0.3 pu237 6.27E-02 52.6 0.2 26.9 2.6 0.3 sm153 6.61E-02 53.1 0.3 27.3 2.6 0.3 sn126 6.59E-02 52.6 0.2 26.9 2.6 0.3 os191 6.88E-02 52.6 0.2 26.9 2.6 0.3 nb95m 7.52E-02 53.0 0.5 27.3 2.6 0.3 rh105 8.17E-02 52.6 0.2 26.9 2.6 0.3 os193 1.08E-01 61.0 3.1 34.4 3.3 0.4 re189 9.10E-02 55.6 0.6 29.5 2.8 0.3 pm149 3.21E-01 83.1 39.9 61.4 9.5 1.2 eu155 9.11E-02 52.6 0.2 26.9 2.6 0.3 gd153 9.29E-02 52.6 0.2 26.9 2.6 0.3 th229 9.61E-02 52.6 0.2 26.9 2.6 0.3 lu177 1.12E-01 52.6 0.2 26.9 2.6 0.3 hf172 1.20E-01 52.6 0.2 26.9 2.6 0.3 ba135m 1.28E-01 52.6 0.2 26.9 2.6 0.3 eu149 1.91E-01 58.6 1.0 32.0 2.9 0.3 yb175 1.69E-01 52.6 0.2 26.9 2.6 0.3 ce144 1.80E-01 52.6 0.2 26.9 2.6 0.3 be7 1.93E-01 52.6 0.2 26.9 2.6 0.3 re186 2.10E-01 54.4 1.4 28.8 2.8 0.3 cr51 1.98E-01 52.6 0.2 26.9 2.6 0.3 xe133m 1.99E-01 52.6 0.2 26.9 2.6 0.3 pa231 2.09E-01 52.7 0.3 27.0 2.6 0.3 ag111 2.47E-01 54.7 1.8 29.0 2.9 0.3 ir192 2.36E-01 52.9 0.5 27.3 2.6 0.3 xe129m 4.35E-01 52.6 0.2 26.9 2.6 0.3 ra224 4.99E-01 52.9 0.3 27.2 2.6 0.3 Robatel Technologies, LLC Page 5-39

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab ac225 5.06E-01 52.9 0.3 27.2 2.6 0.3 kr81 5.11E-01 52.6 0.2 26.9 2.6 0.3 ra226 5.69E-01 52.7 0.3 27.0 2.6 0.3 np237 5.86E-01 52.6 0.2 26.9 2.6 0.3 as77 7.12E-01 61.5 1.4 34.5 3.1 0.3 pt195m 6.57E-01 52.6 0.2 26.9 2.6 0.3 ni59 3.30E+00 78.1 43.3 58.6 9.5 1.2 xe131m 1.02E+00 52.6 0.2 26.9 2.6 0.3 sn113 1.09E+00 52.7 0.3 27.0 2.6 0.3 kr85 2.62E+00 84.9 4.7 54.8 4.5 0.5 th231 1.60E+00 52.6 0.2 26.9 2.6 0.3 dy166 1.70E+00 52.6 0.2 26.9 2.6 0.3 nb91 3.47E+00 84.9 4.7 54.8 4.5 0.5 am243 2.81E+00 53.0 0.4 27.3 2.6 0.3 w188 3.13E+00 52.6 0.2 26.9 2.6 0.3 th228 3.85E+00 52.7 0.3 27.0 2.6 0.3 tb161 4.57E+00 58.6 1.0 32.0 2.9 0.3 es254 4.48E+00 52.6 0.2 26.9 2.6 0.3 u230 6.10E+00 52.7 0.3 27.0 2.6 0.3 la137 1.73E+01 106.0 21.8 79.6 7.1 0.9 te125m 7.07E+00 52.6 0.2 26.9 2.6 0.3 th234 7.08E+00 52.6 0.2 26.9 2.6 0.3 rn222 1.50E+01 84.9 4.7 54.8 4.5 0.5 am242m 1.20E+01 52.6 0.2 26.9 2.6 0.3 te127m 3.73E+01 101.1 18.7 74.2 6.5 0.8 es253 1.90E+01 55.8 2.5 29.0 3.1 0.3 w181 1.72E+01 52.6 0.2 26.9 2.6 0.3 pt193m 1.78E+01 52.6 0.2 26.9 2.6 0.3 po210 7.99E+01 62.2 47.0 53.8 9.5 1.2 u233 2.67E+01 55.1 2.0 28.2 3.0 0.3 u232 2.54E+01 52.7 0.3 27.0 2.6 0.3 xe133 2.72E+01 52.6 0.2 26.9 2.6 0.3 th230 2.84E+01 52.6 0.3 26.9 2.6 0.3 am241 6.02E+01 59.3 4.2 33.7 3.3 0.4 ac227 4.82E+01 52.6 0.2 26.9 2.6 0.3 u234 5.50E+01 52.7 0.3 27.0 2.6 0.3 pd103 6.91E+01 52.6 0.2 26.9 2.6 0.3 th232 7.27E+01 52.6 0.2 26.9 2.6 0.3 cd113m 8.69E+01 52.6 0.2 26.9 2.6 0.3 sr90 3.64E+02 65.5 45.8 31.2 9.5 1.2 u236 9.99E+01 52.6 0.2 26.9 2.6 0.3 v49 1.07E+02 53.8 0.4 27.9 2.7 0.3 pu239 1.14E+02 54.1 1.1 28.4 2.8 0.3 w185 1.07E+02 52.6 0.2 26.9 2.6 0.3 pu236 1.44E+02 57.2 1.7 31.2 2.9 0.3 cf252 1.34E+02 52.6 0.2 26.9 2.6 0.3 Robatel Technologies, LLC Page 5-40

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab cm242 4.05E+02 79.1 18.2 46.2 6.3 0.7 u238 1.86E+02 52.6 0.2 26.9 2.6 0.3 cl36 3.35E+02 84.9 4.7 54.8 4.5 0.5 cm244 8.69E+02 88.5 38.3 64.5 9.5 1.2 pu240 2.60E+02 53.1 0.5 27.4 2.6 0.3 ca41 4.10E+02 52.6 0.2 26.9 2.6 0.3 pu238 1.62E+03 89.9 29.4 59.0 8.1 1.0 sm145 5.73E+02 52.6 0.2 26.9 2.6 0.3 pu242 6.96E+02 52.6 0.2 26.9 2.6 0.3 pm147 7.00E+02 52.6 0.2 26.9 2.6 0.3 er169 1.53E+03 52.6 0.2 26.9 2.6 0.3 fe55 1.63E+03 52.6 0.2 26.9 2.6 0.3 pu241 3.67E+03 52.6 0.2 26.9 2.6 0.3 bi210 1.96E+04 52.6 0.2 26.9 2.6 0.3 bk249 5.28E+04 52.6 0.2 26.9 2.6 0.3 pr143 2.15E+05 87.3 45.3 79.1 9.5 1.3 tc97 5.81E+05 52.6 0.2 26.9 2.6 0.3 ca45 5.83E+05 52.6 0.2 26.9 2.6 0.3 ge71 5.83E+05 52.6 0.2 26.9 2.6 0.3 nb93m 5.83E+05 52.6 0.2 26.9 2.6 0.3 mo93 5.83E+05 52.6 0.2 26.9 2.6 0.3 tc97m 5.83E+05 52.6 0.2 26.9 2.6 0.3 cd109 5.83E+05 52.6 0.2 26.9 2.6 0.3 sn113m 5.83E+05 52.6 0.2 26.9 2.6 0.3 sn119m 5.83E+05 52.6 0.2 26.9 2.6 0.3 sn121m 5.83E+05 52.6 0.2 26.9 2.6 0.3 te123 5.83E+05 52.6 0.2 26.9 2.6 0.3 i125 5.83E+05 52.6 0.2 26.9 2.6 0.3 i129 5.83E+05 52.6 0.2 26.9 2.6 0.3 cs131 5.83E+05 52.6 0.2 26.9 2.6 0.3 pm145 5.83E+05 52.6 0.2 26.9 2.6 0.3 sm151 5.83E+05 52.6 0.2 26.9 2.6 0.3 tb157 5.83E+05 52.6 0.2 26.9 2.6 0.3 dy159 5.83E+05 52.6 0.2 26.9 2.6 0.3 tm170 5.83E+05 52.6 0.2 26.9 2.6 0.3 tm171 5.83E+05 52.6 0.2 26.9 2.6 0.3 os194 5.83E+05 52.6 0.2 26.9 2.6 0.3 pt193 5.83E+05 52.6 0.2 26.9 2.6 0.3 tl204 5.83E+05 52.6 0.2 26.9 2.6 0.3 pb205 5.83E+05 52.6 0.2 26.9 2.6 0.3 pb210 5.83E+05 52.6 0.2 26.9 2.6 0.3 ra225 5.83E+05 52.6 0.2 26.9 2.6 0.3 ra228 5.83E+05 52.6 0.2 26.9 2.6 0.3 np235 5.83E+05 52.6 0.2 26.9 2.6 0.3 cm246 5.83E+05 52.6 0.2 26.9 2.6 0.3 cm248 5.83E+05 52.6 0.2 26.9 2.6 0.3 Robatel Technologies, LLC Page 5-41

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-5 NCT Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Vehicle Source Strength 2 Meter from Occupied Radionuclide Density Package Surface (mrem/hr) Edge of Vehicle Position (Ci/g) (mrem/hr)

(mrem/hr)

X Side Bottom Top Side Cab cf250 5.83E+05 52.6 0.2 26.9 2.6 0.3 se72 5.83E+05 52.6 0.2 26.9 2.6 0.3 as73 5.83E+05 52.6 0.2 26.9 2.6 0.3 te118 5.83E+05 52.6 0.2 26.9 2.6 0.3 sb119 5.83E+05 52.6 0.2 26.9 2.6 0.3 nd140 5.83E+05 52.6 0.2 26.9 2.6 0.3 yb166 5.83E+05 52.6 0.2 26.9 2.6 0.3 h3 5.83E+05 52.6 0.2 26.9 2.6 0.3 ni63 5.83E+05 52.6 0.2 26.9 2.6 0.3 sr89 5.83E+05 52.6 0.2 26.9 2.6 0.3 tc99 5.83E+05 52.6 0.2 26.9 2.6 0.3 am242 5.83E+05 52.6 0.2 26.9 2.6 0.3 c14 5.83E+05 52.6 0.2 26.9 2.6 0.3 Robatel Technologies, LLC Page 5-42

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top na24 3.38E-06 34.7 53.9 53.0 bi208 3.76E-06 33.4 53.7 53.1 cs144 5.27E-06 34.5 53.2 55.0 y88 1.02E-05 39.6 54.7 48.3 la140 1.63E-05 42.1 53.8 49.5 sb124 2.19E-05 42.4 54.5 48.3 eu156 2.26E-05 39.2 53.7 50.5 sc48 2.67E-05 60.2 58.7 58.9 la138 3.87E-05 48.2 57.0 51.4 tb156 4.10E-05 53.0 56.9 53.4 ag106m 4.45E-05 63.2 56.2 55.8 lu169 4.55E-05 48.4 56.0 52.3 na22 5.38E-05 62.5 59.8 57.2 sb120m 5.51E-05 76.6 60.8 64.9 i124 5.54E-05 45.6 54.0 49.9 br82 5.93E-05 71.3 57.8 59.0 lu172 6.68E-05 69.0 59.2 60.6 ta182 6.93E-05 62.0 59.4 58.7 ca47 7.23E-05 56.1 60.1 56.6 sc46 7.66E-05 82.3 60.2 67.2 te131m 7.96E-05 61.5 56.0 57.3 eu152 8.35E-05 60.6 58.2 56.3 as72 8.49E-05 63.9 54.5 59.4 tm172 9.32E-05 45.6 55.9 50.2 eu154 9.69E-05 68.0 60.0 60.0 cs136 1.09E-04 93.7 63.3 71.1 pm148 1.10E-04 51.2 57.2 52.5 ge69 1.25E-04 65.2 57.1 58.4 tb160 1.56E-04 86.0 61.1 68.3 sn125 1.82E-04 48.3 55.2 54.8 rh102m 1.88E-04 121.6 61.1 74.0 gd147 1.93E-04 93.3 59.0 65.5 tc96 2.23E-04 185.0 66.3 101.6 sr83 2.25E-04 69.2 55.0 56.2 k40 2.42E-04 46.0 56.9 50.7 co60 2.50E-04 70.0 97.6 81.7 as76 2.84E-04 54.5 54.9 53.4 nb92 3.41E-04 159.4 67.6 94.4 np238 3.49E-04 102.6 65.8 77.8 am240 4.09E-04 152.4 68.5 95.5 tb158 4.32E-04 128.3 66.4 87.1 pm148m 4.79E-04 229.9 64.0 98.9 ag110m 4.84E-04 84.9 93.4 80.5 ho166m 5.45E-04 250.0 66.0 110.6 pa232 5.96E-04 173.8 68.6 100.2 sb126 6.14E-04 372.3 62.5 127.2 Robatel Technologies, LLC Page 5-43

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top nb94 6.40E-04 304.1 72.3 140.0 rb84 6.41E-04 167.5 63.1 90.3 tm168 6.65E-04 231.9 67.1 108.8 pr144 6.75E-04 36.0 52.6 49.7 fe59 7.11E-04 82.1 100.4 88.5 rh99 7.83E-04 100.7 55.2 58.5 tm165 8.21E-04 118.2 58.9 68.0 zr89 1.18E-03 80.2 53.6 53.0 rh102 1.23E-03 123.5 56.3 64.9 as71 1.37E-03 194.4 58.7 75.4 ru106 1.57E-03 96.4 55.1 61.7 tc95m 1.59E-03 281.8 67.1 115.2 rb86 1.82E-03 88.0 64.7 70.9 tc98 1.97E-03 663.3 68.0 208.2 ag108m 2.12E-03 733.4 66.3 209.9 ho166 2.17E-03 44.3 55.6 50.5 ir194m 2.25E-03 855.9 52.2 168.6 zn65 2.34E-03 104.8 103.1 102.3 cs132 2.34E-03 359.9 55.3 112.9 lu171 2.56E-03 390.4 68.8 151.2 nb95 2.59E-03 526.8 72.8 190.2 zr95 2.62E-03 526.8 72.8 190.2 tb153 2.83E-03 220.5 62.5 88.5 te121m 3.15E-03 186.3 57.2 69.7 ag105 3.32E-03 311.8 61.6 96.4 sb127 3.61E-03 469.1 60.1 138.4 sb122 3.62E-03 277.3 54.9 86.2 pm144 3.14E-03 950.0 43.6 213.8 cd115m 4.17E-03 73.8 61.8 64.6 kr79 4.27E-03 228.3 56.6 81.3 pm146 4.35E-03 644.9 69.8 194.6 gd149 4.88E-03 524.1 64.0 139.7 i126 5.10E-03 428.5 55.8 118.3 cs134 6.03E-03 263.9 104.3 138.8 os185 6.57E-03 266.8 72.1 130.4 cu64 6.61E-03 238.9 52.8 77.0 pm143 6.71E-03 526.8 72.8 190.2 co58 6.96E-03 218.0 100.1 128.3 as74 7.27E-03 747.3 43.2 141.1 ba131 7.28E-03 529.7 57.6 104.4 br77 9.13E-03 538.8 58.1 127.3 ce143 9.23E-03 458.4 59.6 121.5 pm151 1.27E-02 814.1 60.7 171.7 mn54 1.65E-02 483.9 115.9 235.7 y91 1.80E-02 54.9 60.0 56.3 sb125 1.29E-02 950.0 30.4 175.2 Robatel Technologies, LLC Page 5-44

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top sr85 1.15E-02 950.0 8.6 140.9 rb83 1.23E-02 950.0 11.5 145.8 te121 1.15E-02 950.0 8.4 140.3 sn123 2.46E-02 87.8 64.6 70.8 i131 1.71E-02 950.0 17.6 121.7 ag110 3.68E-02 268.4 54.5 94.6 cm241 1.16E-02 950.0 2.6 82.0 lu176 1.12E-02 950.0 0.9 75.7 au198 2.01E-02 950.0 15.0 92.1 se75 1.14E-02 950.0 0.9 75.7 y87 1.16E-02 950.0 0.9 76.0 rh101 1.30E-02 950.0 0.9 75.7 in114m 4.87E-02 945.3 60.9 208.4 hf181 1.36E-02 950.0 1.1 76.5 ru103 1.83E-02 950.0 7.8 99.8 cs129 2.97E-02 950.0 20.0 104.8 xe127 1.73E-02 950.0 0.9 75.8 bi210m 2.20E-02 950.0 6.3 94.6 ru97 2.01E-02 950.0 2.9 80.4 yb169 1.97E-02 950.0 0.9 75.7 co57 1.99E-02 950.0 1.1 76.4 ta183 2.04E-02 950.0 0.9 75.7 np239 2.04E-02 950.0 0.9 75.7 ba133 2.04E-02 950.0 0.9 75.7 zr88 2.05E-02 950.0 0.9 75.7 zn72 2.06E-02 950.0 0.9 75.7 rh101m 2.23E-02 950.0 1.5 81.3 cd115 3.40E-02 950.0 7.2 129.6 pt191 3.21E-02 950.0 8.1 114.7 te132 2.17E-02 950.0 0.9 75.7 cf249 2.18E-02 950.0 1.0 75.9 ba140 3.43E-02 950.0 6.5 124.0 hf182 2.22E-02 950.0 0.9 75.7 tc99m 2.24E-02 950.0 0.9 75.7 sn117m 2.25E-02 950.0 0.9 75.7 te129m 8.36E-02 540.0 62.5 165.0 hf175 2.30E-02 950.0 0.9 75.7 u235 2.32E-02 950.0 0.9 75.7 te123m 2.37E-02 950.0 0.9 75.7 hg203 2.45E-02 950.0 0.9 75.7 cm247 2.45E-02 950.0 0.9 75.7 cf251 2.48E-02 950.0 0.9 75.7 ce139 2.50E-02 950.0 0.9 75.7 pu246 2.54E-02 950.0 0.9 75.7 mo99 9.03E-03 885.2 60.8 186.6 ar37 4.49E-02 950.0 12.2 115.2 Robatel Technologies, LLC Page 5-45

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top sc47 2.93E-02 950.0 0.9 75.7 in113m 3.11E-02 950.0 0.9 75.7 u237 3.13E-02 950.0 0.9 75.7 cm243 3.16E-02 950.0 0.9 75.7 pa233 3.22E-02 950.0 0.9 75.7 ga67 4.31E-02 950.0 7.1 85.3 tb155 3.46E-02 950.0 1.1 76.5 er172 3.72E-02 950.0 1.1 76.7 nd147 6.37E-02 950.0 9.9 136.4 lu177m 3.73E-02 950.0 0.9 75.7 cm245 3.76E-02 950.0 0.9 75.7 np236 3.88E-02 950.0 0.9 75.7 cs137 7.95E-02 950.0 31.3 214.4 cu67 4.00E-02 950.0 0.9 75.7 au199 4.09E-02 950.0 0.9 75.7 ce141 4.14E-02 950.0 0.9 75.7 tm167 4.47E-02 950.0 1.4 79.9 th227 5.64E-02 950.0 1.7 76.9 ra223 5.77E-02 950.0 1.3 77.3 pu237 6.27E-02 950.0 0.9 75.7 sm153 6.61E-02 950.0 1.1 76.7 sn126 6.59E-02 950.0 0.9 75.7 os191 6.88E-02 950.0 0.9 75.7 nb95m 7.52E-02 950.0 1.3 76.7 rh105 8.17E-02 950.0 0.9 75.7 os193 1.08E-01 950.0 5.4 93.0 re189 9.10E-02 950.0 1.6 81.7 pm149 3.21E-01 716.4 61.5 149.1 eu155 9.11E-02 950.0 0.9 75.7 gd153 9.29E-02 950.0 0.9 75.7 th229 9.61E-02 950.0 0.9 75.7 lu177 1.12E-01 950.0 0.9 75.7 hf172 1.20E-01 950.0 0.9 75.7 ba135m 1.28E-01 950.0 0.9 75.7 eu149 1.91E-01 950.0 2.3 87.7 yb175 1.69E-01 950.0 0.9 75.7 ce144 1.80E-01 950.0 0.9 75.7 be7 1.93E-01 950.0 0.9 75.7 re186 2.10E-01 950.0 2.8 80.0 cr51 1.98E-01 950.0 0.9 75.7 xe133m 1.99E-01 950.0 0.9 75.7 pa231 2.09E-01 950.0 0.9 75.8 ag111 2.47E-01 950.0 3.3 80.3 ir192 2.36E-01 950.0 1.3 76.4 xe129m 4.35E-01 950.0 0.9 75.7 ra224 4.99E-01 950.0 1.1 76.3 Robatel Technologies, LLC Page 5-46

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top ac225 5.06E-01 950.0 1.0 76.3 kr81 5.11E-01 950.0 0.9 75.7 ra226 5.69E-01 950.0 0.9 75.8 np237 5.86E-01 950.0 0.9 75.7 as77 7.12E-01 950.0 2.9 93.5 pt195m 6.57E-01 950.0 0.9 75.7 ni59 3.30E+00 408.4 66.1 136.6 xe131m 1.02E+00 950.0 0.9 75.7 sn113 1.09E+00 950.0 1.0 75.9 kr85 2.62E+00 950.0 8.5 140.8 th231 1.60E+00 950.0 0.9 75.7 dy166 1.70E+00 950.0 0.9 75.7 nb91 3.47E+00 950.0 8.5 140.8 am243 2.81E+00 950.0 1.2 76.6 w188 3.13E+00 950.0 0.9 75.7 th228 3.85E+00 950.0 1.0 75.8 tb161 4.57E+00 950.0 2.3 87.7 es254 4.48E+00 950.0 0.9 75.7 u230 6.10E+00 950.0 0.9 75.8 la137 1.73E+01 950.0 37.2 197.8 te125m 7.07E+00 950.0 0.9 75.7 th234 7.08E+00 950.0 0.9 75.7 rn222 1.50E+01 950.0 8.5 140.8 am242m 1.20E+01 950.0 0.9 75.7 te127m 3.73E+01 950.0 32.0 185.3 es253 1.90E+01 950.0 4.2 80.2 w181 1.72E+01 950.0 0.9 75.7 pt193m 1.78E+01 950.0 0.9 75.7 po210 7.99E+01 237.9 72.2 124.1 u233 2.67E+01 950.0 3.4 78.1 u232 2.54E+01 950.0 1.0 75.8 xe133 2.72E+01 950.0 0.9 75.7 th230 2.84E+01 950.0 0.9 75.7 am241 6.02E+01 950.0 7.3 91.3 ac227 4.82E+01 950.0 0.9 75.7 u234 5.50E+01 950.0 0.9 75.8 pd103 6.91E+01 950.0 0.9 75.7 th232 7.27E+01 950.0 0.9 75.7 cd113m 8.69E+01 950.0 0.9 75.7 sr90 3.64E+02 34.1 52.4 49.4 u236 9.99E+01 950.0 0.9 75.7 v49 1.07E+02 950.0 1.2 78.1 pu239 1.14E+02 950.0 2.3 79.0 w185 1.07E+02 950.0 0.9 75.7 pu236 1.44E+02 950.0 3.4 85.6 cf252 1.34E+02 950.0 0.9 75.7 Robatel Technologies, LLC Page 5-47

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top cm242 4.05E+02 950.0 27.5 116.7 u238 1.86E+02 950.0 0.9 75.7 cl36 3.35E+02 950.0 8.5 140.8 cm244 8.69E+02 797.4 59.4 157.5 pu240 2.60E+02 950.0 1.3 76.8 ca41 4.10E+02 950.0 0.9 75.7 pu238 1.62E+03 950.0 45.3 146.1 sm145 5.73E+02 950.0 0.9 75.7 pu242 6.96E+02 950.0 0.9 75.7 pm147 7.00E+02 950.0 0.9 75.7 er169 1.53E+03 950.0 0.9 75.7 fe55 1.63E+03 950.0 0.9 75.7 pu241 3.67E+03 950.0 0.9 75.7 bi210 1.96E+04 950.0 0.9 75.7 bk249 5.28E+04 950.0 0.9 75.7 pr143 2.15E+05 526.8 72.8 190.2 tc97 5.81E+05 950.0 0.9 75.7 ca45 5.83E+05 950.0 0.9 75.7 ge71 5.83E+05 950.0 0.9 75.7 nb93m 5.83E+05 950.0 0.9 75.7 mo93 5.83E+05 950.0 0.9 75.7 tc97m 5.83E+05 950.0 0.9 75.7 cd109 5.83E+05 950.0 0.9 75.7 sn113m 5.83E+05 950.0 0.9 75.7 sn119m 5.83E+05 950.0 0.9 75.7 sn121m 5.83E+05 950.0 0.9 75.7 te123 5.83E+05 950.0 0.9 75.7 i125 5.83E+05 950.0 0.9 75.7 i129 5.83E+05 950.0 0.9 75.7 cs131 5.83E+05 950.0 0.9 75.7 pm145 5.83E+05 950.0 0.9 75.7 sm151 5.83E+05 950.0 0.9 75.7 tb157 5.83E+05 950.0 0.9 75.7 dy159 5.83E+05 950.0 0.9 75.7 tm170 5.83E+05 950.0 0.9 75.7 tm171 5.83E+05 950.0 0.9 75.7 os194 5.83E+05 950.0 0.9 75.7 pt193 5.83E+05 950.0 0.9 75.7 tl204 5.83E+05 950.0 0.9 75.7 pb205 5.83E+05 950.0 0.9 75.7 pb210 5.83E+05 950.0 0.9 75.7 ra225 5.83E+05 950.0 0.9 75.7 ra228 5.83E+05 950.0 0.9 75.7 np235 5.83E+05 950.0 0.9 75.7 cm246 5.83E+05 950.0 0.9 75.7 cm248 5.83E+05 950.0 0.9 75.7 Robatel Technologies, LLC Page 5-48

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.4.4-6 HAC Gamma Dose Rates for the Maximum Radionuclide Loading (Cont.)

Source Strength 1 Meter from Package Surface Radionuclide Density (mrem/hr)

(Ci/g)

X Side Bottom Top cf250 5.83E+05 950.0 0.9 75.7 se72 5.83E+05 950.0 0.9 75.7 as73 5.83E+05 950.0 0.9 75.7 te118 5.83E+05 950.0 0.9 75.7 sb119 5.83E+05 950.0 0.9 75.7 nd140 5.83E+05 950.0 0.9 75.7 yb166 5.83E+05 950.0 0.9 75.7 h3 5.83E+05 950.0 0.9 75.7 ni63 5.83E+05 950.0 0.9 75.7 sr89 5.83E+05 950.0 0.9 75.7 tc99 5.83E+05 950.0 0.9 75.7 am242 5.83E+05 950.0 0.9 75.7 c14 5.83E+05 950.0 0.9 75.7 5.5 Shielding Evaluation - Mass Restricted Resins and Filters This mass restricted filter shielding evaluation is an extension of the dose rate calculations outlined in Section 5.4. The packaging geometry, materials, and all assumptions for the effects of NCT and HAC are retained, with the sole change being an alternative approach to the contents modeling.

This supplemental evaluation calculates dose rates for content volumes equivalent to 500, 1,000, and 1,500 lbs of radioactive filters, with the goal of allowing increased radionuclide specific activity limits by implementing a mass restriction on the total quantity of radioactive contents.

This allows for flexibility in filter contents by permitting lighter loads with higher specific activities or heavier loads with lower specific activities. The resulting radionuclide specific activity limits are set for four total filter content mass bands: 500 lbs, 1,000 lbs, 1,500 lbs, or 15,000 lbs, with the 15,000 lbs limits based on the full cavity contents outlined in Section 5.4.

For typical filter contents, the primary dose rate contribution is almost exclusively from Co-60, with small contributions from the other 7 radionuclides analyzed explicitly in Section 5.4. All other isotopes analyzed using the generic energy line method outlined in Section 5.4 have a minimal or negligible contribution to external dose rates. As a result, the increased specific activity limits determined in this analysis are only for these 8 explicitly analyzed radionuclides. All radionuclides other than the 8 explicitly analyzed default to the bounding generic energy line method specific activity limits, based on a cavity filled entirely by the radioactive contents but no self-shielding credited (i.e., contents modeled as void in MCNP).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.5.6 Specific Activity Limits Each MCNP case tallies the photon flux () at the dose rate location of interest, with two multipliers applied in a dose rate conversion function () and a per Curie multiplier. The dose rate conversion function is based on the ANSI/ANS-6.1.1-1977 Flux-to-Dose conversion factors in mrem/hr (see Table 5.4.3-1), applied using the DE and DF cards in MCNP. The per Curie multiplier is applied using the FM card on each tally and is simply a factor of 3.7E10 Bq/Ci.

Noting that MCNP tallies are normalized per emitted particle, the output values (R) are:

mrem mrem/hr disintegration 2 disintegration/s R[ ] = [ cm ] 3.7E10 [ ] [ hr

]

Ci emitted emitted Ci cm2 s To account for the statistical uncertainty in the calculated dose rate 2 is added to each, based on the calculated fractional standard deviation from MCNP. Each radionuclide has a characteristic number of gammas emitted per disintegration (I). This factor is applied to give dose rate per Curie values (D) for each radionuclide:

mrem/hr mrem/hr disintegration emitted D[ ] = R[ ] I[ ]

Ci Ci emitted disintegration The corresponding activity limit is determined based on the calculated dose rate and the regulatory dose rate limit (D) at each location of interest reduced by 5% (e.g., 190 mrem/hr for the package surface). The overall activity limit is set as the minimum activity limit determined from all regulatory locations (x):

mrem D [ ]

ALimit [Ci] = minx { hr }

mrem/hr D[ Ci ]

Finally, the specific activity limit in Ci/g is calculated by dividing the determined activity limit by the content mass (m) of interest (i.e., 500, 1,000, or 1,500 lbs) converted to grams:

Ci ALimit [Ci]

aLimit [ ] =

g m [g]

The maximum dose rate per Curie values (D) at each regulatory dose rate location for each content mass along with the corresponding activity limit (ALimit) are presented in Table 5.5.6-1 through Table 5.5.6-8 for each radionuclide analyzed. The specific activity limits (aLimit) for each radionuclide and content mass are presented in Table 5.5.7-1.

A dose rate plot from the results in the NCT side mesh tallies for Cs-137 and Co-60 are shown in and Figure 5.5.6-1 and Figure 5.5.6-2, respectively. Both figures are for the case resulting in the set activity limit for the given radionuclide, with cases shown in Figure 5.5.5-1. Due to its lower gamma emission energy (Eg=0.66 MeV), the maximum case for Cs-137, as shown in Figure 5.5.6-1, is with the contents concentrated in the top corner of the cask cavity, which maximizes streaming over the lead shield. Due to its higher gamma energies (Eg=1.17 & 1.33 MeV). Figure 5.5.6-2 shows that the peak Co-60 dose rates are from the case where the contents are stretched along the entire height of the cask at a slightly reduced density to minimize self-shielding from the radioactive contents. These dose rate profiles show the relatively uniform dose rate gradient in the cell tally regions and no potential streaming effects that could result in an underpredicted maximum dose rate for any radionuclide.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.5.6-1 Ag-110m Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 5.221E-01 363.9 NCT 2-meter 4.174E-02 227.6 500 227.6 NCT Cab 4.911E-03 386.9 HAC 1-meter 3.094E-01 3,070.6 NCT Surface 2.744E-01 692.5 NCT 2-meter 3.013E-02 315.3 1000 315.3 NCT Cab 3.828E-03 496.4 HAC 1-meter 1.748E-01 5,435.4 NCT Surface 1.853E-01 1,025.6 NCT 2-meter 2.382E-02 398.8 1500 398.8 NCT Cab 3.085E-03 615.8 HAC 1-meter 1.321E-01 7,190.6 Table 5.5.6-2 Cs-137 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 4.554E-03 41,724.3 NCT 2-meter 2.006E-04 47,351.2 500 41,724.3 NCT Cab 2.073E-05 91,649.9 HAC 1-meter 1.496E-02 63,510.9 NCT Surface 2.383E-03 79,723.8 NCT 2-meter 1.157E-04 82,102.7 1000 79,723.8 NCT Cab 1.329E-05 142,974.1 HAC 1-meter 8.537E-03 111,285.7 NCT Surface 1.606E-03 118,280.9 NCT 2-meter 8.314E-05 114,268.5 1500 114,268.5 NCT Cab 1.076E-05 176,652.6 HAC 1-meter 6.911E-03 137,456.8 Table 5.5.6-3 Co-58 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 3.524E-02 5,391.4 NCT 2-meter 2.808E-03 3,383.4 500 3,383.4 NCT Cab 4.660E-04 4,076.9 HAC 1-meter 5.103E-02 18,616.6 NCT Surface 1.824E-02 10,415.8 NCT 2-meter 2.040E-03 4,656.5 1000 4,656.5 NCT Cab 3.255E-04 5,837.8 HAC 1-meter 2.772E-02 34,276.4 NCT Surface 1.256E-02 15,124.5 NCT 2-meter 1.675E-03 5,671.1 1500 5,671.1 NCT Cab 2.976E-04 6,385.4 HAC 1-meter 2.040E-02 46,577.4 Robatel Technologies, LLC Page 5-56

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.5.6-4 Co-60 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 1.012E+00 187.7 NCT 2-meter 8.080E-02 117.6 500 117.6 NCT Cab 9.010E-03 210.9 HAC 1-meter 4.833E-01 1,965.7 NCT Surface 5.236E-01 362.9 NCT 2-meter 5.779E-02 164.4 1000 164.4 NCT Cab 6.675E-03 284.6 HAC 1-meter 3.164E-01 3,002.4 NCT Surface 3.516E-01 540.5 NCT 2-meter 4.469E-02 212.6 1500 212.6 NCT Cab 5.265E-03 360.9 HAC 1-meter 2.530E-01 3,754.4 Table 5.5.6-5 Cs-134 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 4.098E-02 4,636.8 NCT 2-meter 3.461E-03 2,744.7 500 2,744.7 NCT Cab 4.111E-04 4,621.6 HAC 1-meter 7.338E-02 12,945.7 NCT Surface 2.110E-02 9,006.8 NCT 2-meter 2.443E-03 3,889.2 1000 3,889.2 NCT Cab 3.077E-04 6,174.2 HAC 1-meter 4.338E-02 21,898.5 NCT Surface 1.413E-02 13,442.6 NCT 2-meter 1.887E-03 5,033.5 1500 5,033.5 NCT Cab 2.336E-04 8,133.8 HAC 1-meter 3.195E-02 29,730.4 Table 5.5.6-6 Fe-59 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 3.656E-01 519.7 NCT 2-meter 2.991E-02 317.6 500 317.6 NCT Cab 3.497E-03 543.3 HAC 1-meter 1.999E-01 4,752.0 NCT Surface 1.908E-01 996.0 NCT 2-meter 2.107E-02 450.8 1000 450.8 NCT Cab 2.567E-03 740.1 HAC 1-meter 1.196E-01 7,943.4 NCT Surface 1.275E-01 1,489.8 NCT 2-meter 1.624E-02 585.1 1500 585.1 NCT Cab 2.021E-03 939.9 HAC 1-meter 9.194E-02 10,332.8 Robatel Technologies, LLC Page 5-57

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.5.6-7 Mn-54 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 1.902E-02 9,989.2 NCT 2-meter 1.414E-03 6,716.4 500 6,716.4 NCT Cab 1.786E-04 10,636.8 HAC 1-meter 4.120E-02 23,060.6 NCT Surface 1.002E-02 18,958.1 NCT 2-meter 9.699E-04 9,794.8 1000 9,794.8 NCT Cab 1.248E-04 15,221.0 HAC 1-meter 2.590E-02 36,685.2 NCT Surface 6.803E-03 27,929.2 NCT 2-meter 7.281E-04 13,048.0 1500 13,048.0 NCT Cab 9.629E-05 19,732.3 HAC 1-meter 1.989E-02 47,767.63 Table 5.5.6-8 Zn-65 Dose Rates and Activity Limits Mass (lbs) Location DR/Ci (mrem/hr/Ci) ALimit (Ci) Overall ALimit (Ci)

NCT Surface 1.115E-01 1,704.4 NCT 2-meter 9.227E-03 1,029.5 500 1,029.5 NCT Cab 1.109E-03 1,714.0 HAC 1-meter 7.471E-02 12,715.3 NCT Surface 5.788E-02 3,282.5 NCT 2-meter 6.540E-03 1,452.5 1000 1,452.5 NCT Cab 7.830E-04 2,426.6 HAC 1-meter 4.692E-02 20,246.2 NCT Surface 3.835E-02 4,953.8 NCT 2-meter 4.971E-03 1,911.3 1500 1,911.3 NCT Cab 6.495E-04 2,925.1 HAC 1-meter 3.101E-02 30,634.0 Robatel Technologies, LLC Page 5-58

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.5.7 Dose Rate Compliance A summary of the specific activity limits for each content mass and radionuclide is provided in Table 5.5.7-1. The values in this table are calculated by dividing the overall activity limits for each radionuclide in Table 5.5.6-1 through Table 5.5.6-8 by the respective filter content mass in grams.

For example, the 500 lbs specific-activity limit for Co-60 is calculated as:

117.6 [Ci] 4 Ci g = 5.18 10 [ g ]

500 [lb] 453.59 [ ]

lb Note that the 15,000 lb specific activity limits provided in Table 5.5.7-1 are taken from Table 5.7.3-1, based on the original analysis where the entire cavity is filled with radioactive contents.

The 15,000 lb mass limit is based on the total content mass limit for the RT-100 package, so no radioactive content mass can ever reach this limit as a secondary container is always required.

The specific activity limits are used for the demonstration of compliance with regulatory dose rate limits by calculating the sum of the fractions (SOF) based on the maximum specific activities of all filter contents. The listed mass-based specific activity limits for the eight radionuclides in Table 5.5.7-1 can be used for radioactive filter contents when the total mass of the filter contents is less than the set mass limit. The dose rate contribution from all other radionuclides is minimal, so, to ensure a conservative demonstration of compliance with external dose rates, the bounding full cavity specific activity limits listed in Table 5.7.3-1 can be used for any radionuclide not included in Table 5.5.7-1.

n ai [Ci/g]

SOF =

aLimit,i [Ci/g]

i=1 Note that the specific activity for each radionuclide (ai) is selected as the maximum measured specific activity for the respective radionuclide across all filters in the contents. If there is any rounding of the sum of the fractions value, it should always be conservatively rounded up. A sum of the fractions less than or equal to 1 demonstrates that no external dose rates will exceed the regulatory limit and the contents are permissible for shipment.

Multiple conservatisms are applied in this analysis to ensure dose rates are not exceeded when the calculated specific activity limits are utilized, including:

1. A 5% margin is applied directly to the regulatory limits.

2. No secondary container is considered in the analysis. This neglects the additional spacing and shielding provided by the secondary container.

3. Only the low-density portion of the filters is considered in the contents. All filters have a steel frame, which would otherwise provide significant gamma self-shielding in the content materials, that is not credited.

4. The filter contents are modeled solely as activated carbon, which results in the least photon attenuation of four materials examined in Section 5.4.4.2. The filter contents are modeled in the worst-case configuration and at the worst-case material density in the package cavity for each radionuclide individually.

5. The maximum specific activity for each radionuclide in the contents is applied for the sum of the fractions calculation, effectively neglecting any activity variation and assuming the peak specific activity through all radioactive contents.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.5.7-1 Mass Restricted Filter Specific Activity Limit Summary Specific Activity Limit (Ci/g)

Radionuclide 500 lbs 1,000 lbs 1,500 lbs 15,000 lbs Ag-110m 1.004E-03 6.952E-04 5.861E-04 4.840E-04 Cs-137 1.840E-01 1.758E-01 1.679E-01 7.950E-02 Co-58 1.492E-02 1.027E-02 8.335E-03 6.960E-03 Co-60 5.184E-04 3.624E-04 3.124E-04 2.500E-04 Cs-134 1.210E-02 8.574E-03 7.398E-03 6.030E-03 Fe-59 1.400E-03 9.939E-04 8.600E-04 7.110E-04 Mn-54 2.961E-02 2.159E-02 1.918E-02 1.650E-02 Zn-65 4.540E-03 3.202E-03 2.809E-03 2.340E-03 5.6 Shielding Evaluation - Activated Hardware The activated hardware or activated metal (terms used interchangeably) shielding evaluation is similar to the mass restricted filter analysis outlined in Section 5.5, except the content material analyzed is activated hardware. This is including but not limited to fuel channels, velocity limiters, reactor internals, as well as other general hardware materials that have undergone neutron activation. The only change for the MCNP dose rate calculations in this analysis is the content modeling.

Similar to the mass restricted filter analysis, in this analysis, the permissible radioactive contents are restricted by mass, while still assuming the maximum activity throughout. This allows for variable specific activity limits for the activated hardware contents in the RT-100, while ensuring calculated dose rates are still bounding. Providing multiple mass limits allows for smaller quantities of material with higher specific activities or larger quantities of material with lower specific activities. The imposed mass limits are 1,000, 2,000, 8,000, and 13,000 total lbs. of hardware contents. The lower mass limits are set based on a review of typical mass limits from previous hardware content shipments. The 13,000 lb. limit is set as an upper limit for the package.

Although the maximum content mass for the RT-100 cask is 15,000 lbs., a secondary container will always be utilized, and shoring may be present. As a result, the maximum realistic quantity of activated hardware expected in the RT-100 will be less than the 13,000 lb. limit.

For typical hardware contents, the primary dose rate contribution is almost exclusively from Co-60, with small contributions from the other 7 radionuclides analyzed explicitly in Section 5.4 and Section 5.5. All other isotopes analyzed using the generic energy line method outlined in Section 5.4 have a minimal or negligible contribution to external dose rates. As a result, specific activity limits determined in this analysis are only for the 8 explicitly analyzed radionuclides. All radionuclides other than the 8 explicitly analyzed default to the bounding generic energy line method specific activity limits, based on a cavity filled entirely by the radioactive contents but no self-shielding credited (i.e., contents modeled as void in MCNP).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b) 5.6.6 Specific Activity Limits As with the mass restricted filter analysis discussed in Section 5.5.6, each MCNP case tallies the photon flux () at the dose rate location of interest, with two multipliers applied in a dose rate conversion function () and a per Curie multiplier. The dose rate conversion function is based on the ANSI/ANS-6.1.1-1977 Flux-to-Dose conversion factors in mrem/hr (see Table 5.4.3-1),

applied using the DE and DF cards in MCNP. The per Curie multiplier is applied using the FM card on each tally and is simply a factor of 3.7E10 Bq/Ci. Noting that MCNP tallies are normalized per emitted particle, the output values (R) are:

mrem mrem/hr disintegration cm 2 disintegration/s R[ ] = [ ] 3.7E10 [ ] [ hr

]

Ci emitted emitted Ci cm2 s To account for the statistical uncertainty in the calculated dose rate, 2 is added to each, based on the calculated fractional standard deviation from MCNP. Each radionuclide has a characteristic number of gammas emitted per disintegration (I). This factor is applied to result in dose rate per Curie values (D) for each radionuclide:

mrem/hr mrem/hr disintegration emitted D[ ] = R[ ] I[ ]

Ci Ci emitted disintegration The corresponding activity limit is determined based on the calculated dose rate and the regulatory dose rate limit (D) at each location of interest reduced by 5% (e.g., 190 mrem/hr for the package surface). The overall activity limit is set as the minimum activity limit determined from all regulatory locations (x):

mrem D [

ALimit [Ci] = minx { hr ] }

mrem/hr D[ Ci ]

Finally, the specific activity limit in Ci/g is calculated by dividing the determined activity limit by the content mass (m) of interest (i.e., 1,000, 2,000, 8,000, or 13,000 lbs) converted to grams:

Ci ALimit [Ci]

aLimit [ ] =

g m [g]

The maximum dose rate per Curie values (D) at each regulatory dose rate location for each radionuclide are listed in Table 5.6.6-1 through Table 5.6.6-4 for each mass limit. Each table includes the dose rates calculated for high-density and low-density hardware, individually. For each dose rate location where there are multiple tallies in the radial direction, the value listed is the maximum between the two (see Figure 5.5.5-3 for reference)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The maximum dose rates at each regulatory location for each radionuclide and mass limit are listed in Table 5.6.6-5 through Table 5.6.6-12. The corresponding activity limit based on each location, as well as the overall activity limit (ALimit) are determined based on the listed dose rate and the regulatory dose rate limit (D) at each location of interest reduced by 5% (e.g. 190 mrem/hr for the package surface).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.6.7 Dose Rate Compliance A summary of the specific activity limits for each content mass and radionuclide is provided in Table 5.6.7-1. The values in this table are calculated by dividing the overall activity limits for each radionuclide in Table 5.6.6-5 through Table 5.6.6-12 by the respective hardware mass in grams.

For example, the 1,000 lbs specific-activity limit for Co-60 is calculated as:

351.0 [Ci] 4 Ci g = 7.73 10 [ g ]

1,000 [lb] 453.59 [ ]

lb The specific activity limits are used for the demonstration of compliance with regulatory dose rate limits by calculating the sum of the fractions based on the maximum specific activities of all activated hardware contents. The activated hardware specific activity limits for the eight radionuclides in Table 5.6.7-1 are used along with the bounding full cavity specific activity limits calculated using the generic energy line method and listed in Table 5.7.3-1 for all other radionuclides, to ensure a conservative demonstration of compliance with external dose rates. As the generic energy line method completely neglected self-shielding from the content materials and rounded all gamma emission energies up to the next analyzed gamma energy, these specific activity limits are conservative by a significant margin. A comparison of the specific activity limit calculated for Co-60 using the generic energy line method with void contents is compared to the mass restricted filter content and activated hardware content limits below to demonstrate this conservatism.

Generic Energy Line Method - 2.566E-05 Ci/g Mass Restricted Filters (Table 5.5.7-1) - 2.500E-04 Ci/g to 5.184E-04 Ci/g Activated Hardware (Table 5.6.7-1) - 2.262E-04 Ci/g to 7.739E-04 Ci/g This comparison shows that neglecting self-shielding from the contents has a significant effect on the permissible specific activity limits. For radionuclides with lower gamma energies, the effect is greater as self-shielding provided by the contents has a greater effect. Thus, using the previously developed specific activity limits listed in Table 5.7.3-1 for all other radionuclides is conservative and acceptable for activated hardware contents. Additionally, because external dose rates from both resin/filter contents and activated hardware contents are always driven by Co-60, and to a lesser extent the other seven explicitly analyzed radionuclides, using the specific activity limits from the generic energy line method for all other radionuclides is not overly restrictive.

As with the resin and filter contents, compliance is demonstrated with the sum of the fractions, using the established activated hardware specific activity limit (aLimit,i) and the maximum specific activity of each radionuclide (ai) in the actual hardware contents.

n ai [Ci/g]

SOF =

aLimit,i [Ci/g]

i=1 When there are multiple content types (i.e., high-density hardware, low-density hardware, and/or filters) in a single package, compliance can be demonstrated by calculating the sum of the fractions for each content type individually, then totaling the individual SOF values for a total sum. If there is any rounding of the sum of the fractions value, it should always be conservatively rounded up.

A total sum of the fractions (SOFTotal) less than or equal to 1 indicates that the external dose rates will not exceed the regulatory limit and the contents are permissible for shipment.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 n

SOFTotal = SOFi i=1 Calculating individual sum of the fraction values for each content effectively determines the contribution to external dose rates from each, without consideration for the shielding and spacing provided by other materials present. For example, in a shipment of filters and hardware together, the two contents will provide additional shielding and spacing for each other from the external dose rate locations. However, calculating the sum of the fractions for each content individually neglects these effects. As a result, calculating and summing individual sum of the fraction values for each content results in a conservative estimate of dose rate contribution and is acceptable for demonstrating compliance with external dose rates.

Multiple conservatisms are applied in this analysis to ensure dose rates are not exceeded when the calculated specific activity limits are utilized, including:

1. A 5% margin is applied directly to the regulatory limits.

2. No secondary container is considered in the analysis. This neglects the additional spacing and shielding provided by the secondary container, which for activated hardware contents is typically composed of steel walls.

3. The bounding density of the hardware contents in the low-density or high-density range is considered. For example, activated stainless steels have a density of approximately 8.0 g/cm3, but the bounding contents are analyzed as 7.5 g/cm3. Zirconium alloys have a density of approximately 5.6 g/cm3, but the bounding contents are analyzed as 2.0 g/cm3.

This results in lower, conservative specific activity limits.

4. The contents are modeled solely as the bounding element for the respective density range (i.e., Zr for low-density and Cr for high-density), which provides the lowest gamma attenuation of any material in each density range of the hardware contents.

5. The maximum specific activity for each radionuclide in the contents is applied for the sum of the fractions calculation, effectively neglecting any activity variation and assuming the peak specific activity through all radioactive contents.

Table 5.6.7-1 Activate Hardware Specific Activity Limit Summary Specific Activity Limits (Ci/g)

Radionuclide High-Density Hardware Low-Density Hardware (7.5 9.0 g/cm3) (2.0 < 7.5 g/cm3) 1,000 lbs 2,000 lbs 8,000 lbs 13,000 lbs 1,000 lbs 2,000 lbs 8,000 lbs 13,000 lbs Ag-110m 1.472E-03 1.057E-03 5.715E-04 5.047E-04 8.552E-04 6.236E-04 4.393E-04 4.223E-04 Cs-137 1.697E-01 1.590E-01 1.202E-01 1.062E-01 1.640E-01 1.577E-01 1.163E-01 1.096E-01 Co-58 2.166E-02 1.549E-02 8.565E-03 7.607E-03 1.295E-02 9.091E-03 6.358E-03 6.327E-03 Co-60 7.739E-04 5.485E-04 3.108E-04 2.644E-04 4.463E-04 3.227E-04 2.305E-04 2.262E-04 Cs-134 1.818E-02 1.335E-02 7.122E-03 6.174E-03 1.035E-02 7.784E-03 5.158E-03 5.207E-03 Fe-59 2.167E-03 1.533E-03 8.715E-04 7.489E-04 1.229E-03 8.970E-04 6.447E-04 6.242E-04 Mn-54 4.056E-02 3.586E-02 2.018E-02 1.734E-02 2.807E-02 2.097E-02 1.545E-02 1.530E-02 Zn-65 7.116E-03 5.103E-03 2.785E-03 2.438E-03 4.023E-03 2.926E-03 2.115E-03 2.095E-03 Robatel Technologies, LLC Page 5-78

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.7 Appendix 5.7.1 List of Gamma Radionuclides with Greater than 1 Day Half Life Table 5.7.1-1 List of Gamma Radionuclides with Greater Than 1 Day Half Life Nuclide ID Branch Daughter Yield Nuclide ID Branch Daughter Yield Nuclide ID Branch Daughter Yield na22 110220 cs134 551340 pa232 912320 na24 110240 cs136 551360 pa233 912330 cl36 170360 ba131 561310 u230 922300 ar37 180370 ba133 561330 u232 922320 k40 190400 ba135m 561351 u233 922330 ca41 200410 cs137 561371 br ba137m 0.95 u234 922340 ca45 200450 ba140 561400 u235 922350 ca47 200470 la137 571370 u236 922360 sc46 210460 la138 571380 u237 922370 sc47 210470 la140 571400 u238 922380 sc48 210480 ce139 581390 np235 932350 v49 230490 ce141 581410 np236 932360 cr51 240510 ce143 581430 np237 932370 mn54 250540 ce144 581440 np238 932380 fe55 260550 pr143 591430 np239 932390 fe59 260590 pr144 591440 pu236 942360 co57 270570 nd147 601470 pu237 942370 co58 270580 pm145 611450 pu238 942380 co60 270600 pm147 611470 pu239 942390 ni59 280590 pm148 611480 pu240 942400 cu64 290640 pm148m 611481 pu241 942410 cu67 290670 pm149 611490 pu242 942420 zn65 300650 pm151 611510 pu246 942460 ge71 320710 sm145 621450 am240 952400 as76 330760 sm151 621510 am241 952410 as77 330770 sm153 621530 am242m 952421 se75 340750 eu152 631520 am243 952430 br82 350820 eu154 631540 cm241 962410 kr79 360790 eu155 631550 cm242 962420 kr81 360810 eu156 631560 cm243 962430 kr85 360850 gd153 641530 cm244 962440 rb86 370860 tb157 651570 cm245 962450 sr85 380850 tb160 651600 cm246 962460 sr90 390900 br y90 1 tb161 651610 cm247 962470 y91 390910 dy159 661590 cm248 962480 zr89 400890 dy166 661660 bk249 972490 zr95 400950 ho166 671660 cf249 982490 nb91 410910 ho166m 671661 cf250 982500 nb92 410920 er169 681690 cf251 982510 nb93m 410931 er172 681720 cf252 982520 nb94 410940 tm170 691700 es253 992530 nb95 410950 tm171 691710 es254 992540 nb95m 410951 tm172 691720 be7 40070 mo93 420930 yb169 701690 ga67 310670 tc97 430970 yb175 701750 ge69 320690 tc97m 430971 lu176 711760 as71 330710 tc98 430980 lu177 711770 zn72 300720 mo99 430991 br tc99m 0.88 lu177m 711771 as72 330720 ru97 440970 hf175 721750 se72 340720 ru103 441030 hf181 721810 as73 330730 Robatel Technologies, LLC Page 5-79

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.1-1 (continued)

Nuclide ID Branch Daughter Yield Nuclide ID Branch Daughter Yield Nuclide ID Branch Daughter Yield rh102 451020 hf182 721820 as74 330740 rh105 451050 ta182 731820 br77 350770 ru106 451060 br rh106 1 ta183 731830 rb83 370830 pd103 461030 w181 741810 sr83 380830 ag108m 471081 w185 741850 rb84 370840 ag110 471100 w188 741880 y87 390870 ag110m 471101 re186 751860 y88 390880 ag111 471110 re189 751890 zr88 400880 cd109 481090 os185 761850 tc95m 430951 cd113m 481131 os191 761910 tc96 430960 cd115 481150 os193 761930 rh99 450990 cd115m 481151 os194 761940 rh101 451010 in113m 491131 ir192 771920 rh101m 451011 in114m 491141 ir194m 771941 rh102m 451021 sn113 501130 pt191 781910 ag105 471050 sn113m 501131 pt193 781930 ag106m 471061 sn117m 501171 pt193m 781931 te118 521180 sn119m 501191 pt195m 781951 sb119 511190 sn121m 501211 au198 791980 sb120m 511201 sn123 501230 au199 791990 i124 531240 sn125 501250 hg203 802030 sn126 501260 sb122 511220 tl204 812040 sb127 511270 sb124 511240 pb205 822050 cs129 551290 sb125 511250 bi208 832080 te132 521320 sb126 511260 bi210 832100 nd140 601400 te121 521210 bi210m 832101 pm143 611430 te121m 521211 po210 842100 cs144 551440 te123 521230 pb210 822100 pm144 611440 te123m 521231 rn222 862220 pm146 611460 te125m 521251 ra223 882230 gd147 641470 te127m 521271 ra224 882240 eu149 631490 te129m 521291 ra225 882250 gd149 641490 te131m 521311 ra226 882260 tb153 651530 i125 531250 ra228 882280 tb155 651550 i126 531260 ac225 892250 tb156 651560 i129 531290 ac227 892270 tb158 651580 i131 531310 th227 902270 tm165 691650 xe127 541270 th228 902280 yb166 701660 xe129m 541291 th229 902290 tm167 691670 xe131m 541311 th230 902300 tm168 691680 xe133 541330 th231 902310 lu169 711690 xe133m 541331 th232 902320 lu171 711710 cs131 551310 th234 902340 lu172 711720 cs132 551320 pa231 912310 hf172 721720 Robatel Technologies, LLC Page 5-80

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.7.2 Gamma Radionuclide Responses Table 5.7.2-1 NCT Gamma Dose Rate Responses (mrem/hr/Ci)

NCT Surface Normally NCT 2 Meter Radionuclide 71.47(b)(1) or (2) occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top na24 4.24E+00 3.01E+00 2.16E+00 6.11E-01 1.51E-01 bi208 3.81E+00 2.73E+00 1.98E+00 5.50E-01 1.36E-01 cs144 2.69E+00 1.95E+00 1.45E+00 3.92E-01 9.64E-02 y88 1.43E+00 9.45E-01 6.21E-01 2.03E-01 4.99E-02 la140 8.89E-01 5.82E-01 3.96E-01 1.27E-01 3.11E-02 sb124 6.64E-01 4.32E-01 2.96E-01 9.43E-02 2.29E-02 eu156 6.34E-01 4.34E-01 2.97E-01 9.15E-02 2.22E-02 sc48 5.39E-01 3.54E-01 2.51E-01 7.74E-02 1.88E-02 la138 3.76E-01 2.40E-01 1.66E-01 5.34E-02 1.30E-02 tb156 3.56E-01 2.32E-01 1.61E-01 5.04E-02 1.23E-02 ag106m 3.28E-01 2.11E-01 1.51E-01 4.65E-02 1.14E-02 lu169 3.18E-01 2.08E-01 1.44E-01 4.54E-02 1.10E-02 na22 2.76E-01 1.78E-01 1.28E-01 3.84E-02 9.44E-03 sb120m 2.63E-01 1.73E-01 1.32E-01 3.75E-02 9.17E-03 i124 2.63E-01 1.73E-01 1.20E-01 3.73E-02 9.05E-03 br82 2.47E-01 1.58E-01 1.16E-01 3.48E-02 8.48E-03 lu172 2.16E-01 1.44E-01 1.07E-01 3.09E-02 7.59E-03 ta182 2.11E-01 1.37E-01 9.87E-02 2.98E-02 7.26E-03 ca47 2.04E-01 1.32E-01 9.46E-02 2.86E-02 7.02E-03 sc46 1.88E-01 1.24E-01 9.61E-02 2.70E-02 6.57E-03 te131m 1.82E-01 1.22E-01 8.86E-02 2.60E-02 6.31E-03 eu152 1.74E-01 1.12E-01 8.06E-02 2.47E-02 6.03E-03 as72 1.69E-01 1.18E-01 8.96E-02 2.43E-02 5.97E-03 tm172 1.56E-01 1.00E-01 6.79E-02 2.22E-02 5.38E-03 eu154 1.52E-01 9.87E-02 7.29E-02 2.13E-02 5.23E-03 pm148 1.33E-01 8.48E-02 5.94E-02 1.88E-02 4.60E-03 cs136 1.32E-01 8.93E-02 7.23E-02 1.89E-02 4.69E-03 ge69 1.16E-01 7.55E-02 5.44E-02 1.65E-02 4.00E-03 tb160 9.28E-02 6.15E-02 4.79E-02 1.33E-02 3.23E-03 rh102m 8.02E-02 5.06E-02 4.27E-02 1.10E-02 2.70E-03 sn125 7.86E-02 5.44E-02 3.87E-02 1.14E-02 2.77E-03 gd147 7.59E-02 4.93E-02 3.82E-02 1.07E-02 2.61E-03 sr83 6.53E-02 4.24E-02 3.09E-02 9.20E-03 2.24E-03 tc96 6.48E-02 4.41E-02 4.52E-02 9.27E-03 2.30E-03 k40 6.02E-02 3.84E-02 2.64E-02 8.55E-03 2.09E-03 co601 5.60E-02 3.38E-02 2.65E-02 8.26E-03 1.01E-03 Robatel Technologies, LLC Page 5-81

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally 71.47(b)(1) or (2) NCT 2 Meter Radionuclide occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top as76 5.15E-02 3.42E-02 2.42E-02 7.28E-03 1.77E-03 nb92 4.23E-02 2.91E-02 2.71E-02 6.06E-03 1.48E-03 np238 4.09E-02 2.84E-02 2.39E-02 5.92E-03 1.47E-03 am240 3.44E-02 2.45E-02 2.29E-02 5.05E-03 1.24E-03 tb158 3.30E-02 2.29E-02 2.03E-02 4.79E-03 1.17E-03 sb126 2.94E-02 1.49E-02 2.00E-02 3.37E-03 8.33E-04 ag110m1 2.90E-02 1.74E-02 1.38E-02 4.27E-03 5.24E-04 ho166m 2.79E-02 1.78E-02 1.99E-02 3.79E-03 9.43E-04 pa232 2.37E-02 1.68E-02 1.64E-02 3.47E-03 8.54E-04 pm148m 3.36E-02 1.99E-02 2.10E-02 4.31E-03 1.07E-03 nb94 2.32E-02 1.58E-02 2.04E-02 3.23E-03 8.18E-04 rb84 2.24E-02 1.53E-02 1.46E-02 3.22E-03 8.00E-04 tm168 2.21E-02 1.48E-02 1.60E-02 3.11E-03 7.73E-04 pr144 2.12E-02 1.47E-02 1.00E-02 3.06E-03 7.43E-04 fe591 1.97E-02 1.19E-02 9.70E-03 2.91E-03 3.57E-04 rh99 1.92E-02 1.19E-02 8.93E-03 2.64E-03 6.42E-04 tm165 1.80E-02 1.14E-02 9.10E-03 2.52E-03 6.11E-04 zr89 1.29E-02 7.86E-03 5.62E-03 1.76E-03 4.33E-04 rh102 1.24E-02 7.53E-03 5.93E-03 1.68E-03 4.07E-04 as71 1.16E-02 6.76E-03 5.99E-03 1.51E-03 3.70E-04 tc98 1.10E-02 4.62E-03 9.49E-03 1.05E-03 2.69E-04 ag108m 1.05E-02 4.21E-03 8.88E-03 9.77E-04 2.49E-04 ru106 9.55E-03 6.08E-03 4.73E-03 1.31E-03 3.21E-04 tc95m 9.44E-03 6.12E-03 7.01E-03 1.30E-03 3.24E-04 ir194m 1.06E-02 3.32E-03 6.86E-03 9.18E-04 2.23E-04 cs132 8.10E-03 3.69E-03 4.78E-03 8.84E-04 2.15E-04 rb86 7.93E-03 5.43E-03 4.39E-03 1.14E-03 2.85E-04 pm144 7.84E-03 1.79E-03 6.00E-03 5.43E-04 1.34E-04 nb95 7.33E-03 3.81E-03 6.64E-03 7.98E-04 2.07E-04 zr95 7.25E-03 3.76E-03 6.57E-03 7.88E-04 2.05E-04 lu171 6.84E-03 3.80E-03 5.52E-03 8.08E-04 2.05E-04 ho166 6.68E-03 4.30E-03 2.86E-03 9.54E-04 2.31E-04 zn651 5.98E-03 3.65E-03 3.27E-03 8.84E-04 1.09E-04 sb127 5.38E-03 2.44E-03 3.63E-03 5.73E-04 1.41E-04 tb153 5.40E-03 3.35E-03 3.17E-03 7.30E-04 1.78E-04 ag105 5.01E-03 2.79E-03 2.97E-03 6.22E-04 1.55E-04 te121m 4.92E-03 2.89E-03 2.37E-03 6.56E-04 1.59E-04 sb122 4.82E-03 2.40E-03 2.49E-03 5.71E-04 1.38E-04 Robatel Technologies, LLC Page 5-82

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally 71.47(b)(1) or (2) NCT 2 Meter Radionuclide occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top pm146 4.61E-03 2.17E-03 4.01E-03 4.75E-04 1.22E-04 kr79 3.80E-03 2.10E-03 1.97E-03 4.84E-04 1.17E-04 i126 3.75E-03 1.68E-03 2.29E-03 4.05E-04 9.91E-05 gd149 3.71E-03 1.87E-03 2.63E-03 4.24E-04 1.05E-04 cd115m 3.49E-03 2.32E-03 1.76E-03 4.96E-04 1.21E-04 as74 3.20E-03 9.61E-04 1.84E-03 2.84E-04 6.68E-05 pm143 2.83E-03 1.47E-03 2.56E-03 3.08E-04 7.99E-05 cu64 2.54E-03 1.30E-03 1.22E-03 3.12E-04 7.49E-05 ba131 2.47E-03 1.19E-03 1.40E-03 2.84E-04 6.94E-05 cs1341 2.25E-03 1.34E-03 1.61E-03 3.43E-04 4.36E-05 os185 2.15E-03 1.55E-03 1.86E-03 3.14E-04 7.92E-05 co581 2.01E-03 1.18E-03 1.35E-03 2.97E-04 3.83E-05 ce143 1.95E-03 9.61E-04 1.25E-03 2.24E-04 5.49E-05 sb125 1.62E-03 3.00E-04 1.18E-03 1.05E-04 2.54E-05 pm151 1.62E-03 6.63E-04 1.20E-03 1.62E-04 4.00E-05 sr85 1.61E-03 9.07E-05 1.04E-03 8.62E-05 1.89E-05 te121 1.60E-03 8.81E-05 1.03E-03 8.52E-05 1.86E-05 br77 2.06E-03 9.35E-04 1.33E-03 2.26E-04 5.55E-05 rb83 1.53E-03 1.16E-04 1.01E-03 8.54E-05 1.90E-05 mn541 1.05E-03 5.09E-04 9.71E-04 1.25E-04 1.65E-05 cm241 1.04E-03 2.41E-05 5.56E-04 5.29E-05 1.13E-05 lu176 1.02E-03 4.84E-06 5.22E-04 5.03E-05 1.05E-05 se75 1.00E-03 4.83E-06 5.14E-04 4.95E-05 1.03E-05 tc99m 5.12E-04 2.42E-06 2.62E-04 2.52E-05 5.27E-06 y87 9.93E-04 5.00E-06 5.09E-04 4.89E-05 1.02E-05 i131 9.19E-04 1.33E-04 5.96E-04 5.84E-05 1.36E-05 rh101 8.83E-04 4.18E-06 4.52E-04 4.35E-05 9.09E-06 mo99 2.41E-03 9.25E-04 1.83E-03 2.29E-04 5.70E-05 hf181 8.50E-04 6.10E-06 4.37E-04 4.20E-05 8.81E-06 y91 8.21E-04 5.33E-04 3.79E-04 1.15E-04 2.82E-05 ru103 7.53E-04 5.15E-05 4.44E-04 4.12E-05 9.16E-06 au198 7.21E-04 1.09E-04 3.83E-04 4.98E-05 1.13E-05 xe127 6.63E-04 3.25E-06 3.39E-04 3.26E-05 6.83E-06 bi210m 6.05E-04 3.45E-05 3.47E-04 3.25E-05 7.15E-06 ru97 5.93E-04 1.67E-05 3.14E-04 3.09E-05 6.59E-06 sn123 5.85E-04 4.01E-04 3.24E-04 8.39E-05 2.11E-05 yb169 5.81E-04 2.79E-06 2.97E-04 2.86E-05 5.98E-06 co57 5.78E-04 4.02E-06 2.97E-04 2.86E-05 5.99E-06 Robatel Technologies, LLC Page 5-83

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally 71.47(b)(1) or (2) NCT 2 Meter Radionuclide occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top ta183 5.61E-04 2.66E-06 2.87E-04 2.76E-05 5.78E-06 np239 5.60E-04 2.65E-06 2.87E-04 2.76E-05 5.77E-06 ba133 5.60E-04 2.65E-06 2.87E-04 2.76E-05 5.77E-06 zr88 5.58E-04 2.64E-06 2.86E-04 2.75E-05 5.75E-06 zn72 5.55E-04 2.63E-06 2.84E-04 2.73E-05 5.71E-06 rh101m 5.41E-04 6.06E-06 2.86E-04 2.70E-05 5.67E-06 te132 5.28E-04 2.50E-06 2.70E-04 2.60E-05 5.44E-06 cf249 5.25E-04 2.91E-06 2.69E-04 2.59E-05 5.42E-06 cs129 5.24E-04 9.56E-05 2.95E-04 3.84E-05 8.74E-06 hf182 5.15E-04 2.44E-06 2.64E-04 2.54E-05 5.31E-06 sn117m 5.08E-04 2.41E-06 2.60E-04 2.50E-05 5.23E-06 cd115 5.08E-04 2.53E-05 3.19E-04 2.69E-05 5.85E-06 hf175 4.98E-04 2.36E-06 2.55E-04 2.45E-05 5.13E-06 u235 4.93E-04 2.37E-06 2.52E-04 2.43E-05 5.08E-06 ba140 4.86E-04 2.24E-05 3.02E-04 2.56E-05 5.55E-06 pt191 4.85E-04 3.11E-05 2.95E-04 2.65E-05 5.84E-06 te123m 4.82E-04 2.29E-06 2.47E-04 2.38E-05 4.97E-06 ag110 4.79E-04 2.45E-04 2.72E-04 5.61E-05 1.37E-05 in114m 4.73E-04 1.68E-04 3.76E-04 4.24E-05 1.06E-05 hg203 4.68E-04 2.22E-06 2.39E-04 2.30E-05 4.81E-06 cm247 4.67E-04 2.21E-06 2.39E-04 2.30E-05 4.80E-06 cf251 4.61E-04 2.18E-06 2.36E-04 2.27E-05 4.75E-06 ce139 4.59E-04 2.17E-06 2.35E-04 2.26E-05 4.72E-06 pu246 4.51E-04 2.14E-06 2.31E-04 2.22E-05 4.64E-06 sc47 3.90E-04 1.85E-06 2.00E-04 1.92E-05 4.02E-06 in113m 3.69E-04 1.75E-06 1.89E-04 1.81E-05 3.79E-06 u237 3.66E-04 1.73E-06 1.87E-04 1.80E-05 3.77E-06 cm243 3.62E-04 1.71E-06 1.85E-04 1.78E-05 3.73E-06 pa233 3.55E-04 1.68E-06 1.82E-04 1.75E-05 3.66E-06 ar37 3.41E-04 3.44E-05 2.13E-04 2.00E-05 4.54E-06 tb155 3.33E-04 2.13E-06 1.71E-04 1.65E-05 3.45E-06 er172 3.10E-04 2.32E-06 1.60E-04 1.54E-05 3.22E-06 lu177m 3.07E-04 1.45E-06 1.57E-04 1.51E-05 3.16E-06 cm245 3.04E-04 1.44E-06 1.56E-04 1.50E-05 3.13E-06 np236 2.95E-04 1.40E-06 1.51E-04 1.45E-05 3.04E-06 ga67 2.87E-04 2.16E-05 1.57E-04 1.69E-05 3.71E-06 cu67 2.86E-04 1.36E-06 1.46E-04 1.41E-05 2.95E-06 nd147 2.81E-04 1.90E-05 1.81E-04 1.53E-05 3.38E-06 Robatel Technologies, LLC Page 5-84

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally 71.47(b)(1) or (2) NCT 2 Meter Radionuclide occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top au199 2.80E-04 1.32E-06 1.43E-04 1.38E-05 2.88E-06 ce141 2.77E-04 1.31E-06 1.42E-04 1.36E-05 2.85E-06 tm167 2.67E-04 2.58E-06 1.40E-04 1.32E-05 2.78E-06 cs1371 2.58E-04 2.58E-05 1.80E-04 1.45E-05 1.76E-06 te129m 2.45E-04 1.06E-04 1.84E-04 2.47E-05 6.18E-06 th227 2.05E-04 2.99E-06 1.06E-04 1.04E-05 2.19E-06 ra223 2.01E-04 1.79E-06 1.04E-04 9.98E-06 2.10E-06 pu237 1.83E-04 8.65E-07 9.35E-05 9.00E-06 1.88E-06 sm153 1.75E-04 1.14E-06 8.99E-05 8.63E-06 1.81E-06 sn126 1.74E-04 8.23E-07 8.89E-05 8.56E-06 1.79E-06 os191 1.66E-04 7.89E-07 8.52E-05 8.20E-06 1.71E-06 nb95m 1.53E-04 1.40E-06 7.91E-05 7.63E-06 1.60E-06 rh105 1.40E-04 6.64E-07 7.17E-05 6.90E-06 1.44E-06 re189 1.33E-04 1.54E-06 7.05E-05 6.62E-06 1.39E-06 eu155 1.26E-04 5.96E-07 6.43E-05 6.19E-06 1.29E-06 gd153 1.23E-04 5.84E-07 6.31E-05 6.07E-06 1.27E-06 os193 1.23E-04 6.19E-06 6.92E-05 6.71E-06 1.45E-06 th229 1.19E-04 5.64E-07 6.09E-05 5.86E-06 1.23E-06 lu177 1.03E-04 4.86E-07 5.25E-05 5.05E-06 1.06E-06 hf172 9.56E-05 4.53E-07 4.89E-05 4.71E-06 9.85E-07 ba135m 8.95E-05 4.24E-07 4.58E-05 4.41E-06 9.22E-07 yb175 6.78E-05 3.21E-07 3.47E-05 3.34E-06 6.98E-07 eu149 6.68E-05 1.19E-06 3.65E-05 3.36E-06 7.12E-07 ce144 6.37E-05 3.02E-07 3.26E-05 3.14E-06 6.55E-07 be7 5.94E-05 2.81E-07 3.04E-05 2.92E-06 6.11E-07 cr51 5.79E-05 2.74E-07 2.96E-05 2.85E-06 5.96E-07 xe133m 5.74E-05 2.72E-07 2.94E-05 2.83E-06 5.91E-07 re186 5.64E-05 1.47E-06 2.99E-05 2.91E-06 6.22E-07 pm149 5.62E-05 2.70E-05 4.16E-05 6.43E-06 1.57E-06 pa231 5.47E-05 2.67E-07 2.80E-05 2.70E-06 5.64E-07 ir192 4.88E-05 4.30E-07 2.51E-05 2.43E-06 5.09E-07 ag111 4.81E-05 1.58E-06 2.55E-05 2.54E-06 5.44E-07 xe129m 2.63E-05 1.25E-07 1.35E-05 1.30E-06 2.71E-07 ra224 2.30E-05 1.51E-07 1.18E-05 1.14E-06 2.38E-07 ac225 2.28E-05 1.24E-07 1.17E-05 1.12E-06 2.35E-07 kr81 2.24E-05 1.06E-07 1.15E-05 1.10E-06 2.30E-07 ra226 2.01E-05 9.93E-08 1.03E-05 9.92E-07 2.07E-07 np237 1.95E-05 9.26E-08 1.00E-05 9.63E-07 2.01E-07 Robatel Technologies, LLC Page 5-85

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally NCT 2 Meter Radionuclide 71.47(b)(1) or (2) occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top as77 1.88E-05 4.38E-07 1.05E-05 9.53E-07 2.03E-07 pt195m 1.74E-05 8.25E-08 8.91E-06 8.58E-07 1.79E-07 xe131m 1.13E-05 5.33E-08 5.76E-06 5.54E-07 1.16E-07 sn113 1.05E-05 5.74E-08 5.39E-06 5.18E-07 1.08E-07 th231 7.14E-06 3.38E-08 3.65E-06 3.51E-07 7.35E-08 kr85 7.04E-06 3.90E-07 4.54E-06 3.76E-07 8.21E-08 dy166 6.73E-06 3.19E-08 3.45E-06 3.32E-07 6.93E-08 nb91 5.32E-06 2.95E-07 3.43E-06 2.84E-07 6.21E-08 ni59 5.15E-06 2.86E-06 3.87E-06 6.27E-07 1.56E-07 am243 4.11E-06 3.15E-08 2.12E-06 2.04E-07 4.27E-08 w188 3.66E-06 1.73E-08 1.87E-06 1.80E-07 3.77E-08 th228 2.98E-06 1.70E-08 1.53E-06 1.47E-07 3.08E-08 tb161 2.79E-06 4.99E-08 1.53E-06 1.40E-07 2.97E-08 es254 2.56E-06 1.21E-08 1.31E-06 1.26E-07 2.63E-08 u230 1.88E-06 9.22E-09 9.62E-07 9.26E-08 1.94E-08 te125m 1.62E-06 7.67E-09 8.28E-07 7.97E-08 1.67E-08 th234 1.62E-06 7.66E-09 8.28E-07 7.97E-08 1.67E-08 la137 1.33E-06 2.74E-07 9.99E-07 8.85E-08 2.16E-08 rn222 1.23E-06 6.83E-08 7.96E-07 6.58E-08 1.44E-08 am242m 9.57E-07 4.53E-09 4.90E-07 4.71E-08 9.85E-09 w181 6.67E-07 3.16E-09 3.41E-07 3.29E-08 6.87E-09 pt193m 6.44E-07 3.05E-09 3.30E-07 3.17E-08 6.63E-09 es253 6.40E-07 2.88E-08 3.33E-07 3.50E-08 7.52E-09 te127m 5.89E-07 1.09E-07 4.32E-07 3.81E-08 9.22E-09 u232 4.51E-07 2.46E-09 2.31E-07 2.22E-08 4.65E-09 u233 4.50E-07 1.66E-08 2.30E-07 2.42E-08 5.16E-09 xe133 4.21E-07 1.99E-09 2.15E-07 2.07E-08 4.33E-09 th230 4.03E-07 1.92E-09 2.06E-07 1.99E-08 4.15E-09 ac227 2.38E-07 1.13E-09 1.22E-07 1.17E-08 2.45E-09 am241 2.14E-07 1.52E-08 1.22E-07 1.21E-08 2.68E-09 u234 2.08E-07 1.02E-09 1.07E-07 1.03E-08 2.14E-09 po210 1.69E-07 1.28E-07 1.46E-07 2.59E-08 6.50E-09 pd103 1.66E-07 7.85E-10 8.48E-08 8.16E-09 1.71E-09 th232 1.57E-07 7.46E-10 8.05E-08 7.75E-09 1.62E-09 cd113m 1.32E-07 6.24E-10 6.74E-08 6.49E-09 1.36E-09 u236 1.15E-07 5.43E-10 5.87E-08 5.65E-09 1.18E-09 v49 1.09E-07 8.22E-10 5.66E-08 5.39E-09 1.13E-09 w185 1.07E-07 5.06E-10 5.46E-08 5.26E-09 1.10E-09 Robatel Technologies, LLC Page 5-86

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally NCT 2 Meter Radionuclide 71.47(b)(1) or (2) occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top pu239 1.03E-07 2.08E-09 5.42E-08 5.26E-09 1.12E-09 pu236 8.60E-08 2.57E-09 4.69E-08 4.42E-09 9.48E-10 cf252 8.54E-08 4.04E-10 4.37E-08 4.20E-09 8.79E-10 u238 6.15E-08 2.92E-10 3.15E-08 3.03E-09 6.34E-10 cl36 5.51E-08 3.06E-09 3.56E-08 2.94E-09 6.43E-10 pu240 4.44E-08 4.12E-10 2.29E-08 2.21E-09 4.64E-10 cm242 4.25E-08 9.80E-09 2.48E-08 3.37E-09 7.87E-10 sr90 3.91E-08 2.73E-08 1.86E-08 5.67E-09 1.38E-09 ca41 2.79E-08 1.32E-10 1.43E-08 1.37E-09 2.87E-10 cm244 2.22E-08 9.60E-09 1.61E-08 2.38E-09 5.80E-10 sm145 2.00E-08 9.46E-11 1.02E-08 9.84E-10 2.06E-10 pu242 1.64E-08 7.79E-11 8.41E-09 8.10E-10 1.69E-10 pm147 1.64E-08 7.75E-11 8.37E-09 8.06E-10 1.68E-10 pu238 1.21E-08 3.95E-09 7.93E-09 1.09E-09 2.65E-10 er169 7.46E-09 3.54E-11 3.82E-09 3.68E-10 7.68E-11 fe55 7.02E-09 3.33E-11 3.59E-09 3.46E-10 7.23E-11 pu241 3.12E-09 1.48E-11 1.60E-09 1.54E-10 3.21E-11 bi210 5.83E-10 2.76E-12 2.99E-10 2.87E-11 6.01E-12 bk249 2.17E-10 1.03E-12 1.11E-10 1.07E-11 2.23E-12 pr143 8.82E-11 4.58E-11 7.99E-11 9.59E-12 2.49E-12 tc97 1.97E-11 9.34E-14 1.01E-11 9.71E-13 2.03E-13 ca45 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 ge71 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 nb93m 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 mo93 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 tc97m 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 cd109 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sn113m 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sn119m 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sn121m 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 te123 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 i125 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 i129 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 cs131 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 pm145 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sm151 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 tb157 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 dy159 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 Robatel Technologies, LLC Page 5-87

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-1 (Continued)

NCT Surface Normally NCT 2 Meter Radionuclide 71.47(b)(1) or (2) occupied space 71.47(b)(3) 71.47(b)(4)

Side Bottom Top tm170 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 tm171 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 os194 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 pt193 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 tl204 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 pb205 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 pb210 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 ra225 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 ra228 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 np235 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 cm246 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 cm248 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 cf250 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 se72 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 as73 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 te118 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sb119 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 nd140 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 yb166 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 h3 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 ni63 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 sr89 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 tc99 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 am242 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 c14 1.96E-11 9.30E-14 1.00E-11 9.67E-13 2.02E-13 Note 1: Marked nuclides are analyzed individually in Sections 5.4, 5.5, 5.6. Values displayed in this table are for the general resins and filters shielding evaluation from Section 5.4. For the mass restricted resins/filters and activated hardware gamma responses on these 8 primary isotopes, refer to Sections 5.5 and 5.6, respectively.

Robatel Technologies, LLC Page 5-88

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 HAC Gamma Dose Rate Responses (mrem/hr/Ci)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top na24 2.23E+00 3.47E+00 3.41E+00 bi208 1.93E+00 3.11E+00 3.07E+00 cs144 1.43E+00 2.20E+00 2.27E+00 y88 8.44E-01 1.17E+00 1.03E+00 la140 5.64E-01 7.20E-01 6.62E-01 sb124 4.20E-01 5.41E-01 4.79E-01 eu156 3.77E-01 5.17E-01 4.86E-01 sc48 4.91E-01 4.79E-01 4.80E-01 la138 2.71E-01 3.20E-01 2.89E-01 tb156 2.81E-01 3.02E-01 2.83E-01 ag106m 3.09E-01 2.75E-01 2.73E-01 lu169 2.32E-01 2.68E-01 2.50E-01 na22 2.53E-01 2.42E-01 2.32E-01 sb120m 3.02E-01 2.40E-01 2.56E-01 i124 1.79E-01 2.12E-01 1.96E-01 br82 2.61E-01 2.12E-01 2.16E-01 lu172 2.25E-01 1.93E-01 1.97E-01 ta182 1.94E-01 1.86E-01 1.84E-01 ca47 1.69E-01 1.81E-01 1.70E-01 sc46 2.34E-01 1.71E-01 1.91E-01 te131m 1.68E-01 1.53E-01 1.56E-01 eu152 1.58E-01 1.52E-01 1.47E-01 as72 1.64E-01 1.40E-01 1.52E-01 tm172 1.06E-01 1.30E-01 1.17E-01 eu154 1.53E-01 1.35E-01 1.35E-01 pm148 1.01E-01 1.13E-01 1.04E-01 cs136 1.86E-01 1.26E-01 1.41E-01 ge69 1.13E-01 9.91E-02 1.01E-01 tb160 1.20E-01 8.54E-02 9.54E-02 rh102m 1.41E-01 7.07E-02 8.57E-02 sn125 5.79E-02 6.61E-02 6.56E-02 gd147 1.05E-01 6.65E-02 7.37E-02 sr83 6.71E-02 5.33E-02 5.44E-02 tc96 1.81E-01 6.47E-02 9.92E-02 k40 4.14E-02 5.12E-02 4.56E-02 co601 6.08E-02 8.49E-02 7.11E-02 Robatel Technologies, LLC Page 5-89

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top as76 4.18E-02 4.21E-02 4.10E-02 nb92 1.02E-01 4.31E-02 6.02E-02 np238 6.39E-02 4.10E-02 4.85E-02 am240 8.10E-02 3.64E-02 5.08E-02 tb158 6.46E-02 3.35E-02 4.39E-02 sb126 1.32E-01 2.21E-02 4.51E-02 ag110m1 3.82E-02 4.20E-02 3.62E-02 ho166m 9.97E-02 2.63E-02 4.41E-02 pa232 6.34E-02 2.50E-02 3.65E-02 pm148m 1.04E-01 2.90E-02 4.49E-02 nb94 1.03E-01 2.46E-02 4.76E-02 rb84 5.68E-02 2.14E-02 3.06E-02 tm168 7.58E-02 2.19E-02 3.56E-02 pr144 1.16E-02 1.70E-02 1.60E-02 fe591 2.51E-02 3.07E-02 2.71E-02 rh99 2.80E-02 1.53E-02 1.63E-02 tm165 3.13E-02 1.56E-02 1.80E-02 zr89 1.48E-02 9.89E-03 9.80E-03 rh102 2.19E-02 9.97E-03 1.15E-02 as71 3.09E-02 9.34E-03 1.20E-02 tc98 7.31E-02 7.49E-03 2.30E-02 ag108m 7.54E-02 6.82E-03 2.16E-02 ru106 1.33E-02 7.62E-03 8.54E-03 tc95m 3.84E-02 9.16E-03 1.57E-02 ir194m 8.27E-02 5.04E-03 1.63E-02 cs132 3.35E-02 5.14E-03 1.05E-02 rb86 1.05E-02 7.74E-03 8.49E-03 pm144 6.58E-02 3.02E-03 1.48E-02 nb95 4.42E-02 6.11E-03 1.60E-02 zr95 4.37E-02 6.04E-03 1.58E-02 lu171 3.32E-02 5.86E-03 1.29E-02 ho166 4.45E-03 5.59E-03 5.07E-03 zn651 9.75E-03 9.59E-03 9.52E-03 sb127 2.83E-02 3.62E-03 8.34E-03 tb153 1.69E-02 4.80E-03 6.80E-03 ag105 2.04E-02 4.03E-03 6.31E-03 te121m 1.29E-02 3.95E-03 4.81E-03 sb122 1.67E-02 3.30E-03 5.18E-03 Robatel Technologies, LLC Page 5-90

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top pm146 3.23E-02 3.49E-03 9.73E-03 kr79 1.16E-02 2.89E-03 4.15E-03 i126 1.83E-02 2.38E-03 5.05E-03 gd149 2.34E-02 2.86E-03 6.23E-03 cd115m 3.85E-03 3.22E-03 3.37E-03 as74 2.24E-02 1.29E-03 4.22E-03 pm143 1.71E-02 2.36E-03 6.16E-03 cu64 7.86E-03 1.74E-03 2.53E-03 ba131 1.58E-02 1.72E-03 3.12E-03 cs1341 9.51E-03 3.76E-03 5.00E-03 os185 8.83E-03 2.38E-03 4.31E-03 co581 6.81E-03 3.13E-03 4.01E-03 ce143 1.08E-02 1.40E-03 2.86E-03 sb125 1.60E-02 5.12E-04 2.95E-03 pm151 1.39E-02 1.04E-03 2.93E-03 sr85 1.80E-02 1.63E-04 2.67E-03 te121 1.79E-02 1.59E-04 2.65E-03 br77 1.28E-02 1.38E-03 3.03E-03 rb83 1.68E-02 2.02E-04 2.57E-03 mn541 6.38E-03 1.53E-03 3.11E-03 cm241 1.78E-02 4.96E-05 1.54E-03 lu176 1.84E-02 1.76E-05 1.47E-03 se75 1.81E-02 1.75E-05 1.45E-03 tc99m 9.23E-03 8.84E-06 7.36E-04 y87 1.79E-02 1.76E-05 1.43E-03 i131 1.21E-02 2.24E-04 1.55E-03 rh101 1.59E-02 1.52E-05 1.27E-03 mo99 2.13E-02 1.46E-03 4.49E-03 hf181 1.52E-02 1.81E-05 1.23E-03 y91 6.64E-04 7.26E-04 6.81E-04 ru103 1.13E-02 9.23E-05 1.19E-03 au198 1.03E-02 1.63E-04 9.99E-04 xe127 1.20E-02 1.16E-05 9.53E-04 bi210m 9.41E-03 6.28E-05 9.37E-04 ru97 1.03E-02 3.18E-05 8.71E-04 sn123 7.75E-04 5.71E-04 6.25E-04 yb169 1.05E-02 1.01E-05 8.36E-04 co57 1.04E-02 1.21E-05 8.34E-04 Robatel Technologies, LLC Page 5-91

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top ta183 1.01E-02 9.69E-06 8.07E-04 np239 1.01E-02 9.68E-06 8.06E-04 ba133 1.01E-02 9.68E-06 8.06E-04 zr88 1.01E-02 9.64E-06 8.03E-04 zn72 1.00E-02 9.58E-06 7.98E-04 rh101m 9.28E-03 1.51E-05 7.94E-04 te132 9.53E-03 9.12E-06 7.60E-04 cf249 9.46E-03 9.73E-06 7.56E-04 cs129 6.96E-03 1.46E-04 7.68E-04 hf182 9.30E-03 8.90E-06 7.41E-04 sn117m 9.17E-03 8.78E-06 7.31E-04 cd115 6.07E-03 4.62E-05 8.28E-04 hf175 9.00E-03 8.61E-06 7.17E-04 u235 8.90E-03 8.57E-06 7.09E-04 ba140 6.03E-03 4.12E-05 7.87E-04 pt191 6.43E-03 5.50E-05 7.76E-04 te123m 8.71E-03 8.33E-06 6.94E-04 ag110 1.58E-03 3.22E-04 5.58E-04 in114m 4.22E-03 2.72E-04 9.30E-04 hg203 8.44E-03 8.08E-06 6.73E-04 cm247 8.42E-03 8.06E-06 6.71E-04 cf251 8.32E-03 7.96E-06 6.63E-04 ce139 8.28E-03 7.92E-06 6.60E-04 pu246 8.14E-03 7.79E-06 6.49E-04 sc47 7.04E-03 6.74E-06 5.61E-04 in113m 6.65E-03 6.37E-06 5.30E-04 u237 6.61E-03 6.32E-06 5.26E-04 cm243 6.53E-03 6.25E-06 5.20E-04 pa233 6.41E-03 6.13E-06 5.11E-04 ar37 4.60E-03 5.91E-05 5.58E-04 tb155 5.97E-03 6.66E-06 4.81E-04 er172 5.55E-03 6.71E-06 4.48E-04 lu177m 5.53E-03 5.30E-06 4.41E-04 cm245 5.49E-03 5.26E-06 4.38E-04 np236 5.32E-03 5.09E-06 4.24E-04 ga67 4.79E-03 3.59E-05 4.30E-04 cu67 5.16E-03 4.94E-06 4.11E-04 nd147 3.24E-03 3.40E-05 4.66E-04 Robatel Technologies, LLC Page 5-92

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top au199 5.05E-03 4.83E-06 4.02E-04 ce141 4.99E-03 4.78E-06 3.98E-04 tm167 4.63E-03 6.76E-06 3.89E-04 cs1371 2.60E-03 8.58E-05 5.87E-04 te129m 1.41E-03 1.63E-04 4.29E-04 th227 3.66E-03 6.58E-06 2.97E-04 ra223 3.58E-03 4.84E-06 2.91E-04 pu237 3.30E-03 3.16E-06 2.63E-04 sm153 3.13E-03 3.53E-06 2.52E-04 sn126 3.14E-03 3.00E-06 2.50E-04 os191 3.00E-03 2.88E-06 2.39E-04 nb95m 2.75E-03 3.71E-06 2.22E-04 rh105 2.53E-03 2.42E-06 2.02E-04 re189 2.27E-03 3.79E-06 1.95E-04 eu155 2.27E-03 2.17E-06 1.81E-04 gd153 2.22E-03 2.13E-06 1.77E-04 os193 1.91E-03 1.09E-05 1.87E-04 th229 2.15E-03 2.06E-06 1.71E-04 lu177 1.85E-03 1.77E-06 1.48E-04 hf172 1.73E-03 1.65E-06 1.38E-04 ba135m 1.62E-03 1.55E-06 1.29E-04 yb175 1.22E-03 1.17E-06 9.75E-05 eu149 1.08E-03 2.59E-06 1.00E-04 ce144 1.15E-03 1.10E-06 9.16E-05 be7 1.07E-03 1.03E-06 8.54E-05 cr51 1.04E-03 9.99E-07 8.32E-05 xe133m 1.04E-03 9.92E-07 8.26E-05 re186 9.85E-04 2.89E-06 8.30E-05 pm149 4.85E-04 4.16E-05 1.01E-04 pa231 9.87E-04 9.58E-07 7.87E-05 ir192 8.75E-04 1.16E-06 7.04E-05 ag111 8.35E-04 2.92E-06 7.06E-05 xe129m 4.76E-04 4.55E-07 3.79E-05 ra224 4.14E-04 4.67E-07 3.32E-05 ac225 4.09E-04 4.20E-07 3.28E-05 kr81 4.04E-04 3.87E-07 3.22E-05 ra226 3.63E-04 3.54E-07 2.90E-05 np237 3.53E-04 3.38E-07 2.81E-05 Robatel Technologies, LLC Page 5-93

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top as77 2.90E-04 8.95E-07 2.85E-05 pt195m 3.14E-04 3.01E-07 2.51E-05 xe131m 2.03E-04 1.94E-07 1.62E-05 sn113 1.89E-04 1.94E-07 1.51E-05 th231 1.29E-04 1.23E-07 1.03E-05 kr85 7.87E-05 7.05E-07 1.17E-05 dy166 1.22E-04 1.16E-07 9.69E-06 nb91 5.95E-05 5.33E-07 8.82E-06 ni59 2.69E-05 4.36E-06 9.01E-06 am243 7.36E-05 9.09E-08 5.94E-06 w188 6.60E-05 6.32E-08 5.26E-06 th228 5.37E-05 5.58E-08 4.28E-06 tb161 4.52E-05 1.08E-07 4.18E-06 es254 4.61E-05 4.42E-08 3.68E-06 u230 3.39E-05 3.30E-08 2.70E-06 te125m 2.92E-05 2.80E-08 2.33E-06 th234 2.92E-05 2.79E-08 2.33E-06 la137 1.19E-05 4.67E-07 2.48E-06 rn222 1.38E-05 1.23E-07 2.04E-06 am242m 1.73E-05 1.65E-08 1.38E-06 w181 1.20E-05 1.15E-08 9.60E-07 pt193m 1.16E-05 1.11E-08 9.26E-07 es253 1.09E-05 4.85E-08 9.19E-07 te127m 5.54E-06 1.87E-07 1.08E-06 u232 8.12E-06 8.29E-09 6.49E-07 u233 7.75E-06 2.75E-08 6.37E-07 xe133 7.60E-06 7.27E-09 6.05E-07 th230 7.27E-06 6.98E-09 5.80E-07 ac227 4.29E-06 4.11E-09 3.42E-07 am241 3.43E-06 2.63E-08 3.30E-07 u234 3.75E-06 3.66E-09 3.00E-07 po210 6.48E-07 1.96E-07 3.38E-07 pd103 2.99E-06 2.86E-09 2.38E-07 th232 2.84E-06 2.72E-09 2.26E-07 cd113m 2.38E-06 2.28E-09 1.90E-07 u236 2.07E-06 1.98E-09 1.65E-07 v49 1.92E-06 2.38E-09 1.58E-07 w185 1.93E-06 1.84E-09 1.54E-07 Robatel Technologies, LLC Page 5-94

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top pu239 1.81E-06 4.31E-09 1.51E-07 pu236 1.43E-06 5.06E-09 1.29E-07 cf252 1.54E-06 1.47E-09 1.23E-07 u238 1.11E-06 1.06E-09 8.85E-08 cl36 6.17E-07 5.52E-09 9.15E-08 pu240 7.94E-07 1.09E-09 6.42E-08 cm242 5.11E-07 1.48E-08 6.27E-08 sr90 2.04E-08 3.13E-08 2.95E-08 ca41 5.04E-07 4.82E-10 4.02E-08 cm244 2.00E-07 1.49E-08 3.94E-08 sm145 3.60E-07 3.45E-10 2.87E-08 pu242 2.97E-07 2.84E-10 2.36E-08 pm147 2.95E-07 2.83E-10 2.35E-08 pu238 1.28E-07 6.08E-09 1.96E-08 er169 1.35E-07 1.29E-10 1.07E-08 fe55 1.27E-07 1.21E-10 1.01E-08 pu241 5.63E-08 5.39E-11 4.49E-09 bi210 1.05E-08 1.01E-11 8.39E-10 bk249 3.92E-09 3.75E-12 3.12E-10 pr143 5.32E-10 7.35E-11 1.92E-10 tc97 3.56E-10 3.40E-13 2.83E-11 ca45 3.54E-10 3.39E-13 2.82E-11 ge71 3.54E-10 3.39E-13 2.82E-11 nb93m 3.54E-10 3.39E-13 2.82E-11 mo93 3.54E-10 3.39E-13 2.82E-11 tc97m 3.54E-10 3.39E-13 2.82E-11 cd109 3.54E-10 3.39E-13 2.82E-11 sn113m 3.54E-10 3.39E-13 2.82E-11 sn119m 3.54E-10 3.39E-13 2.82E-11 sn121m 3.54E-10 3.39E-13 2.82E-11 te123 3.54E-10 3.39E-13 2.82E-11 i125 3.54E-10 3.39E-13 2.82E-11 i129 3.54E-10 3.39E-13 2.82E-11 cs131 3.54E-10 3.39E-13 2.82E-11 pm145 3.54E-10 3.39E-13 2.82E-11 sm151 3.54E-10 3.39E-13 2.82E-11 tb157 3.54E-10 3.39E-13 2.82E-11 dy159 3.54E-10 3.39E-13 2.82E-11 Robatel Technologies, LLC Page 5-95

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.2-2 (Continued)

HAC 1 Meter Radionuclide 71.51(a)(2)

Side Bottom Top tm170 3.54E-10 3.39E-13 2.82E-11 tm171 3.54E-10 3.39E-13 2.82E-11 os194 3.54E-10 3.39E-13 2.82E-11 pt193 3.54E-10 3.39E-13 2.82E-11 tl204 3.54E-10 3.39E-13 2.82E-11 pb205 3.54E-10 3.39E-13 2.82E-11 pb210 3.54E-10 3.39E-13 2.82E-11 ra225 3.54E-10 3.39E-13 2.82E-11 ra228 3.54E-10 3.39E-13 2.82E-11 np235 3.54E-10 3.39E-13 2.82E-11 cm246 3.54E-10 3.39E-13 2.82E-11 cm248 3.54E-10 3.39E-13 2.82E-11 cf250 3.54E-10 3.39E-13 2.82E-11 se72 3.54E-10 3.39E-13 2.82E-11 as73 3.54E-10 3.39E-13 2.82E-11 te118 3.54E-10 3.39E-13 2.82E-11 sb119 3.54E-10 3.39E-13 2.82E-11 nd140 3.54E-10 3.39E-13 2.82E-11 yb166 3.54E-10 3.39E-13 2.82E-11 h3 3.54E-10 3.39E-13 2.82E-11 ni63 3.54E-10 3.39E-13 2.82E-11 sr89 3.54E-10 3.39E-13 2.82E-11 tc99 3.54E-10 3.39E-13 2.82E-11 am242 3.54E-10 3.39E-13 2.82E-11 c14 3.54E-10 3.39E-13 2.82E-11 Note 2: Marked nuclides are analyzed individually in Sections 5.4, 5.5, 5.6. Values displayed in this table are for the general resins and filters shielding evaluation from Section 5.4. For the mass restricted resins/filters and activated hardware gamma responses on these 8 primary isotopes, refer to Sections 5.5 and 5.6, respectively.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.7.3 Radionuclide Maximum Ci/g Loading Limits Table 5.7.3-1 Radionuclide Maximum Ci/g Loading Limits based on Gamma Response Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition ac225 5.06E-01 HAC h3 5.83E+05 HAC ac227 4.82E+01 HAC be7 1.93E-01 HAC ag105 3.32E-03 NCT c14 5.83E+05 HAC ag106m 4.45E-05 NCT na22 5.38E-05 NCT ag108m 2.12E-03 NCT na24 3.38E-06 NCT ag110 3.68E-02 NCT cl36 3.35E+02 HAC ag110m¹ 4.84E-04 NCT ar37 4.49E-02 HAC ag111 2.47E-01 HAC k40 2.42E-04 NCT am240 4.09E-04 NCT ca41 4.10E+02 HAC am241 6.02E+01 HAC ca45 5.83E+05 HAC am242 5.83E+05 HAC sc46 7.66E-05 NCT am242m 1.20E+01 HAC ca47 7.23E-05 NCT am243 2.81E+00 HAC sc47 2.93E-02 HAC ar37 4.49E-02 HAC sc48 2.67E-05 NCT as71 1.37E-03 NCT v49 1.07E+02 HAC as72 8.49E-05 NCT cr51 1.98E-01 HAC as73 5.83E+05 HAC mn54¹ 1.65E-02 NCT as74 7.27E-03 NCT fe55 1.63E+03 HAC as76 2.84E-04 NCT co57 1.99E-02 HAC as77 7.12E-01 HAC co58¹ 6.96E-03 NCT au198 2.01E-02 HAC fe59¹ 7.11E-04 NCT au199 4.09E-02 HAC ni59 3.30E+00 NCT ba131 7.28E-03 NCT co60¹ 2.50E-04 NCT ba133 2.04E-02 HAC ni63 5.83E+05 HAC ba135m 1.28E-01 HAC cu64 6.61E-03 NCT ba140 3.43E-02 HAC zn65¹ 2.34E-03 NCT be7 1.93E-01 HAC cu67 4.00E-02 HAC bi208 3.76E-06 NCT ga67 4.31E-02 HAC bi210 1.96E+04 HAC ge69 1.25E-04 NCT bi210m 2.20E-02 HAC as71 1.37E-03 NCT bk249 5.28E+04 HAC ge71 5.83E+05 HAC br77 9.13E-03 NCT as72 8.49E-05 NCT br82 5.93E-05 NCT se72 5.83E+05 HAC c14 5.83E+05 HAC zn72 2.06E-02 HAC ca41 4.10E+02 HAC as73 5.83E+05 HAC ca45 5.83E+05 HAC as74 7.27E-03 NCT Robatel Technologies, LLC Page 5-97

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition ca47 7.23E-05 NCT se75 1.14E-02 HAC cd109 5.83E+05 HAC as76 2.84E-04 NCT cd113m 8.69E+01 HAC as77 7.12E-01 HAC cd115 3.40E-02 HAC br77 9.13E-03 NCT cd115m 4.17E-03 NCT kr79 4.27E-03 NCT ce139 2.50E-02 HAC kr81 5.11E-01 HAC ce141 4.14E-02 HAC br82 5.93E-05 NCT ce143 9.23E-03 NCT rb83 1.23E-02 HAC ce144 1.80E-01 HAC sr83 2.25E-04 NCT cf249 2.18E-02 HAC rb84 6.41E-04 NCT cf250 5.83E+05 HAC kr85 2.62E+00 HAC cf251 2.48E-02 HAC sr85 1.15E-02 HAC cf252 1.34E+02 HAC rb86 1.82E-03 NCT cl36 3.35E+02 HAC y87 1.16E-02 HAC cm241 1.16E-02 HAC y88 1.02E-05 NCT cm242 4.05E+02 HAC zr88 2.05E-02 HAC cm243 3.16E-02 HAC sr89 5.83E+05 HAC cm244 8.69E+02 NCT zr89 1.18E-03 NCT cm245 3.76E-02 HAC sr90 3.64E+02 NCT cm246 5.83E+05 HAC nb91 3.47E+00 HAC cm247 2.45E-02 HAC y91 1.80E-02 NCT cm248 5.83E+05 HAC nb92 3.41E-04 NCT co57 1.99E-02 HAC mo93 5.83E+05 HAC co58¹ 6.96E-03 NCT nb93m 5.83E+05 HAC co60¹ 2.50E-04 NCT nb94 6.40E-04 NCT cr51 1.98E-01 HAC nb95 2.59E-03 NCT cs129 2.97E-02 HAC zr95 2.62E-03 NCT cs131 5.83E+05 HAC nb95m 7.52E-02 HAC cs132 2.34E-03 NCT tc95m 1.59E-03 NCT cs134¹ 6.03E-03 NCT tc96 2.23E-04 NCT cs136 1.09E-04 NCT ru97 2.01E-02 HAC cs137¹ 7.95E-02 HAC tc97 5.81E+05 HAC cs144 5.27E-06 NCT tc97m 5.83E+05 HAC cu64 6.61E-03 NCT tc98 1.97E-03 NCT cu67 4.00E-02 HAC mo99 9.03E-03 NCT dy159 5.83E+05 HAC rh99 7.83E-04 NCT dy166 1.70E+00 HAC tc99 5.83E+05 HAC er169 1.53E+03 HAC tc99m 2.24E-02 HAC Robatel Technologies, LLC Page 5-98

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition er172 3.72E-02 HAC rh101 1.30E-02 HAC es253 1.90E+01 HAC rh101m 2.23E-02 HAC es254 4.48E+00 HAC rh102 1.23E-03 NCT eu149 1.91E-01 HAC rh102m 1.88E-04 NCT eu152 8.35E-05 NCT pd103 6.91E+01 HAC eu154 9.69E-05 NCT ru103 1.83E-02 HAC eu155 9.11E-02 HAC ag105 3.32E-03 NCT eu156 2.26E-05 NCT rh105 8.17E-02 HAC fe55 1.63E+03 HAC ru106 1.57E-03 NCT fe59¹ 7.11E-04 NCT ag106m 4.45E-05 NCT ga67 4.31E-02 HAC ag108m 2.12E-03 NCT gd147 1.93E-04 NCT cd109 5.83E+05 HAC gd149 4.88E-03 NCT ag110 3.68E-02 NCT gd153 9.29E-02 HAC ag110m¹ 4.84E-04 NCT ge69 1.25E-04 NCT ag111 2.47E-01 HAC ge71 5.83E+05 HAC sn113 1.09E+00 HAC h3 5.83E+05 HAC cd113m 8.69E+01 HAC hf172 1.20E-01 HAC in113m 3.11E-02 HAC hf175 2.30E-02 HAC sn113m 5.83E+05 HAC hf181 1.36E-02 HAC in114m 4.87E-02 NCT hf182 2.22E-02 HAC cd115 3.40E-02 HAC hg203 2.45E-02 HAC cd115m 4.17E-03 NCT ho166 2.17E-03 NCT sn117m 2.25E-02 HAC ho166m 5.45E-04 NCT te118 5.83E+05 HAC i124 5.54E-05 NCT sb119 5.83E+05 HAC i125 5.83E+05 HAC sn119m 5.83E+05 HAC i126 5.10E-03 NCT sb120m 5.51E-05 NCT i129 5.83E+05 HAC te121 1.15E-02 HAC i131 1.71E-02 HAC sn121m 5.83E+05 HAC in113m 3.11E-02 HAC te121m 3.15E-03 NCT in114m 4.87E-02 NCT sb122 3.62E-03 NCT ir192 2.36E-01 HAC sn123 2.46E-02 NCT ir194m 2.25E-03 NCT te123 5.83E+05 HAC k40 2.42E-04 NCT te123m 2.37E-02 HAC kr79 4.27E-03 NCT i124 5.54E-05 NCT kr81 5.11E-01 HAC sb124 2.19E-05 NCT kr85 2.62E+00 HAC i125 5.83E+05 HAC la137 1.73E+01 HAC sb125 1.29E-02 HAC Robatel Technologies, LLC Page 5-99

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition la138 3.87E-05 NCT sn125 1.82E-04 NCT la140 1.63E-05 NCT te125m 7.07E+00 HAC lu169 4.55E-05 NCT i126 5.10E-03 NCT lu171 2.56E-03 NCT sb126 6.14E-04 NCT lu172 6.68E-05 NCT sn126 6.59E-02 HAC lu176 1.12E-02 HAC sb127 3.61E-03 NCT lu177 1.12E-01 HAC xe127 1.73E-02 HAC lu177m 3.73E-02 HAC te127m 3.73E+01 HAC mn54¹ 1.65E-02 NCT cs129 2.97E-02 HAC mo93 5.83E+05 HAC i129 5.83E+05 HAC mo99 9.03E-03 NCT te129m 8.36E-02 NCT na22 5.38E-05 NCT xe129m 4.35E-01 HAC na24 3.38E-06 NCT ba131 7.28E-03 NCT nb91 3.47E+00 HAC cs131 5.83E+05 HAC nb92 3.41E-04 NCT i131 1.71E-02 HAC nb93m 5.83E+05 HAC te131m 7.96E-05 NCT nb94 6.40E-04 NCT xe131m 1.02E+00 HAC nb95 2.59E-03 NCT cs132 2.34E-03 NCT nb95m 7.52E-02 HAC te132 2.17E-02 HAC nd140 5.83E+05 HAC ba133 2.04E-02 HAC nd147 6.37E-02 HAC xe133 2.72E+01 HAC ni59 3.30E+00 NCT xe133m 1.99E-01 HAC ni63 5.83E+05 HAC cs134¹ 6.03E-03 NCT np235 5.83E+05 HAC ba135m 1.28E-01 HAC np236 3.88E-02 HAC cs136 1.09E-04 NCT np237 5.86E-01 HAC cs137¹ 7.95E-02 HAC np238 3.49E-04 NCT la137 1.73E+01 HAC np239 2.04E-02 HAC la138 3.87E-05 NCT os185 6.57E-03 NCT ce139 2.50E-02 HAC os191 6.88E-02 HAC ba140 3.43E-02 HAC os193 1.08E-01 HAC la140 1.63E-05 NCT os194 5.83E+05 HAC nd140 5.83E+05 HAC pa231 2.09E-01 HAC ce141 4.14E-02 HAC pa232 5.96E-04 NCT ce143 9.23E-03 NCT pa233 3.22E-02 HAC pm143 6.71E-03 NCT pb205 5.83E+05 HAC pr143 2.15E+05 NCT pb210 5.83E+05 HAC ce144 1.80E-01 HAC pd103 6.91E+01 HAC cs144 5.27E-06 NCT Robatel Technologies, LLC Page 5-100

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition pm143 6.71E-03 NCT pm144 3.14E-03 HAC pm144 3.14E-03 HAC pr144 6.75E-04 NCT pm145 5.83E+05 HAC pm145 5.83E+05 HAC pm146 4.35E-03 NCT sm145 5.73E+02 HAC pm147 7.00E+02 HAC pm146 4.35E-03 NCT pm148 1.10E-04 NCT gd147 1.93E-04 NCT pm148m 4.79E-04 NCT nd147 6.37E-02 HAC pm149 3.21E-01 NCT pm147 7.00E+02 HAC pm151 1.27E-02 NCT pm148 1.10E-04 NCT po210 7.99E+01 NCT pm148m 4.79E-04 NCT pr143 2.15E+05 NCT eu149 1.91E-01 HAC pr144 6.75E-04 NCT gd149 4.88E-03 NCT pt191 3.21E-02 HAC pm149 3.21E-01 NCT pt193 5.83E+05 HAC pm151 1.27E-02 NCT pt193m 1.78E+01 HAC sm151 5.83E+05 HAC pt195m 6.57E-01 HAC eu152 8.35E-05 NCT pu236 1.44E+02 HAC gd153 9.29E-02 HAC pu237 6.27E-02 HAC sm153 6.61E-02 HAC pu238 1.62E+03 HAC tb153 2.83E-03 NCT pu239 1.14E+02 HAC eu154 9.69E-05 NCT pu240 2.60E+02 HAC eu155 9.11E-02 HAC pu241 3.67E+03 HAC tb155 3.46E-02 HAC pu242 6.96E+02 HAC eu156 2.26E-05 NCT pu246 2.54E-02 HAC tb156 4.10E-05 NCT ra223 5.77E-02 HAC tb157 5.83E+05 HAC ra224 4.99E-01 HAC tb158 4.32E-04 NCT ra225 5.83E+05 HAC dy159 5.83E+05 HAC ra226 5.69E-01 HAC tb160 1.56E-04 NCT ra228 5.83E+05 HAC tb161 4.57E+00 HAC rb83 1.23E-02 HAC tm165 8.21E-04 NCT rb84 6.41E-04 NCT dy166 1.70E+00 HAC rb86 1.82E-03 NCT ho166 2.17E-03 NCT re186 2.10E-01 HAC yb166 5.83E+05 HAC re189 9.10E-02 HAC ho166m 5.45E-04 NCT rh101 1.30E-02 HAC tm167 4.47E-02 HAC rh101m 2.23E-02 HAC tm168 6.65E-04 NCT rh102 1.23E-03 NCT er169 1.53E+03 HAC rh102m 1.88E-04 NCT lu169 4.55E-05 NCT Robatel Technologies, LLC Page 5-101

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition rh105 8.17E-02 HAC yb169 1.97E-02 HAC rh99 7.83E-04 NCT tm170 5.83E+05 HAC rn222 1.50E+01 HAC lu171 2.56E-03 NCT ru103 1.83E-02 HAC tm171 5.83E+05 HAC ru106 1.57E-03 NCT er172 3.72E-02 HAC ru97 2.01E-02 HAC hf172 1.20E-01 HAC sb119 5.83E+05 HAC lu172 6.68E-05 NCT sb120m 5.51E-05 NCT tm172 9.32E-05 NCT sb122 3.62E-03 NCT hf175 2.30E-02 HAC sb124 2.19E-05 NCT yb175 1.69E-01 HAC sb125 1.29E-02 HAC lu176 1.12E-02 HAC sb126 6.14E-04 NCT lu177 1.12E-01 HAC sb127 3.61E-03 NCT lu177m 3.73E-02 HAC sc46 7.66E-05 NCT hf181 1.36E-02 HAC sc47 2.93E-02 HAC w181 1.72E+01 HAC sc48 2.67E-05 NCT hf182 2.22E-02 HAC se72 5.83E+05 HAC ta182 6.93E-05 NCT se75 1.14E-02 HAC ta183 2.04E-02 HAC sm145 5.73E+02 HAC os185 6.57E-03 NCT sm151 5.83E+05 HAC w185 1.07E+02 HAC sm153 6.61E-02 HAC re186 2.10E-01 HAC sn113 1.09E+00 HAC w188 3.13E+00 HAC sn113m 5.83E+05 HAC re189 9.10E-02 HAC sn117m 2.25E-02 HAC os191 6.88E-02 HAC sn119m 5.83E+05 HAC pt191 3.21E-02 HAC sn121m 5.83E+05 HAC ir192 2.36E-01 HAC sn123 2.46E-02 NCT os193 1.08E-01 HAC sn125 1.82E-04 NCT pt193 5.83E+05 HAC sn126 6.59E-02 HAC pt193m 1.78E+01 HAC sr83 2.25E-04 NCT os194 5.83E+05 HAC sr85 1.15E-02 HAC ir194m 2.25E-03 NCT sr89 5.83E+05 HAC pt195m 6.57E-01 HAC sr90 3.64E+02 NCT au198 2.01E-02 HAC ta182 6.93E-05 NCT au199 4.09E-02 HAC ta183 2.04E-02 HAC hg203 2.45E-02 HAC tb153 2.83E-03 NCT tl204 5.83E+05 HAC tb155 3.46E-02 HAC pb205 5.83E+05 HAC tb156 4.10E-05 NCT bi208 3.76E-06 NCT Robatel Technologies, LLC Page 5-102

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition tb157 5.83E+05 HAC bi210 1.96E+04 HAC tb158 4.32E-04 NCT pb210 5.83E+05 HAC tb160 1.56E-04 NCT po210 7.99E+01 NCT tb161 4.57E+00 HAC bi210m 2.20E-02 HAC tc95m 1.59E-03 NCT rn222 1.50E+01 HAC tc96 2.23E-04 NCT ra223 5.77E-02 HAC tc97 5.81E+05 HAC ra224 4.99E-01 HAC tc97m 5.83E+05 HAC ac225 5.06E-01 HAC tc98 1.97E-03 NCT ra225 5.83E+05 HAC tc99 5.83E+05 HAC ra226 5.69E-01 HAC tc99m 2.24E-02 HAC ac227 4.82E+01 HAC te118 5.83E+05 HAC th227 5.64E-02 HAC te121 1.15E-02 HAC ra228 5.83E+05 HAC te121m 3.15E-03 NCT th228 3.85E+00 HAC te123 5.83E+05 HAC th229 9.61E-02 HAC te123m 2.37E-02 HAC th230 2.84E+01 HAC te125m 7.07E+00 HAC u230 6.10E+00 HAC te127m 3.73E+01 HAC pa231 2.09E-01 HAC te129m 8.36E-02 NCT th231 1.60E+00 HAC te131m 7.96E-05 NCT pa232 5.96E-04 NCT te132 2.17E-02 HAC th232 7.27E+01 HAC th227 5.64E-02 HAC u232 2.54E+01 HAC th228 3.85E+00 HAC pa233 3.22E-02 HAC th229 9.61E-02 HAC u233 2.67E+01 HAC th230 2.84E+01 HAC th234 7.08E+00 HAC th231 1.60E+00 HAC u234 5.50E+01 HAC th232 7.27E+01 HAC np235 5.83E+05 HAC th234 7.08E+00 HAC u235 2.32E-02 HAC tl204 5.83E+05 HAC np236 3.88E-02 HAC tm165 8.21E-04 NCT pu236 1.44E+02 HAC tm167 4.47E-02 HAC u236 9.99E+01 HAC tm168 6.65E-04 NCT np237 5.86E-01 HAC tm170 5.83E+05 HAC pu237 6.27E-02 HAC tm171 5.83E+05 HAC u237 3.13E-02 HAC tm172 9.32E-05 NCT np238 3.49E-04 NCT u230 6.10E+00 HAC pu238 1.62E+03 HAC u232 2.54E+01 HAC u238 1.86E+02 HAC u233 2.67E+01 HAC np239 2.04E-02 HAC Robatel Technologies, LLC Page 5-103

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 5.7.3-1 (Continued)

Ordered by Alphanumeric Ordered by Atomic Number Radionuclide Max. Ci/g Condition Radionuclide Max. Ci/g Condition u234 5.50E+01 HAC pu239 1.14E+02 HAC u235 2.32E-02 HAC am240 4.09E-04 NCT u236 9.99E+01 HAC pu240 2.60E+02 HAC u237 3.13E-02 HAC am241 6.02E+01 HAC u238 1.86E+02 HAC cm241 1.16E-02 HAC v49 1.07E+02 HAC pu241 3.67E+03 HAC w181 1.72E+01 HAC am242 5.83E+05 HAC w185 1.07E+02 HAC cm242 4.05E+02 HAC w188 3.13E+00 HAC pu242 6.96E+02 HAC xe127 1.73E-02 HAC am242m 1.20E+01 HAC xe129m 4.35E-01 HAC am243 2.81E+00 HAC xe131m 1.02E+00 HAC cm243 3.16E-02 HAC xe133 2.72E+01 HAC cm244 8.69E+02 NCT xe133m 1.99E-01 HAC cm245 3.76E-02 HAC y87 1.16E-02 HAC cm246 5.83E+05 HAC y88 1.02E-05 NCT pu246 2.54E-02 HAC y91 1.80E-02 NCT cm247 2.45E-02 HAC yb166 5.83E+05 HAC cm248 5.83E+05 HAC yb169 1.97E-02 HAC bk249 5.28E+04 HAC yb175 1.69E-01 HAC cf249 2.18E-02 HAC zn65¹ 2.34E-03 NCT cf250 5.83E+05 HAC zn72 2.06E-02 HAC cf251 2.48E-02 HAC zr88 2.05E-02 HAC cf252 1.34E+02 HAC zr89 1.18E-03 NCT es253 1.90E+01 HAC zr95 2.62E-03 NCT es254 4.48E+00 HAC Note 1: Marked nuclides are analyzed individually in Sections 5.4, 5.5, and 5.6. Values displayed in this table are for the general resins and filter shielding evaluation from Section 5.4. For the mass restricted resins/filters and activated hardware specific activities on these 8 primary isotopes, refer to Sections 5.5 and 5.6, respectively.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.7.4 Process Description for Calculating Maximum Allowed Source Strength Density Generic Energy Line Method:

1. MCNP MCNP6 Generic Energy Line Method Output Files Calculated Dose Rate for Each Low energy Mid energy High energy Additional energy Tally Location (0.5-6.0 MeV) (0.7-1.0 MeV) (1.1-6.0 MeV) (7.0-8.0 MeV) rpbnGenericlo rpbnGenericmo rpbnGenericho rpbnGeneric78o rptnGenericlo rptnGenericmo rptnGenericho rptnGeneric78o rpsnGenericlo rpsnGenericmo rpsnGenericho rpsnGeneric78o rpbaGenericlo rpbaGenericmo rpbaGenericho rpbaGeneric78o rptaGenericlo rptaGenericmo rptaGenericho rptaGeneric78o rpsaGenericlo rpsaGenericmo rpsaGenericho rpsaGeneric78o Compile all MCNP6 output Tally results into Excel sheet Tally_Results.xlsx and

2. Compile determine maximum dose rate calculated for each energy line at each dose rate Results location.

Compile all nuclide photon energy lines from Origen-S data library

3. Generate origen.rev02.mpdkxgam and generate energy grouping based on generic energy lines Nuclide Energy in excel sheet nuclide_energy_groups.xlsx. (As shown for Fe-59 in Figure 5.4.1-1)

Grouping

4. Calculate Nuclide Take nuclide energy groups from nuclide_energy_groups.xlsx and calculated energy Dose Rates and line dose rates from Tally_Results.xlsx to calculate nuclide-specific dose rates and Specific Activity specific activity limits in Excel sheet DoseRateCalc.xlsx. (Results shown in Table Limits 5.5.2-1, Table 5.5.2-2, and Table 5.5.3-1)

Individual Nuclides:

1. MCNP Calculated MCNP6 Individual Nuclide Output Files Dose Rate for Each Nuclide Co-60 Ba-137m Zn-65 Fe-59 Location rpbnCo60o rpbnBa137mo rpbnZn65o rpbnFe59o rptnCo60o rptnBa137mo rptnZn65o rptnFe59o rpsnCo60o rpsnBa137mo rpsnZn65o rpsnFe59o rpbaCo60o rpbaBa137mo rpbaZn65o rpbaFe59o rptaCo60o rptaBa137mo rptaZn65o rptaFe59o rpsaCo60o rpsaBa137mo rpsaZn65o rpsaFe59o Co-58 Ag-110m Mn-54 Cs-134 rpbnCo58o rpbnAg110mo rpbnMn54o rpbnCs134o rptnCo58o rptnAg110mo rptnMn54o rptnCs134o rpsnCo58o rpsnAg110mo rpsnMn54o rpsnCs134o rpbaCo58o rpbaAg110mo rpbaMn54o rpbaCs134o rptaCo58o rptaAg110mo rptaMn54o rptaCs134o rpsaCo58o rpsaAg110mo rpsaMn54o rpsaCs134o Compile all MCNP6 output Tally results and calculates dose rates and specific

2. Compile Results activity limits for each nuclide individually in Excel sheets: Co60.xlsx, Co58.xlsx, Cs137.xlsx, Ag110m.xlsx, Zn65.xlsx, Mn54.xlsx, Fe59.xlsx, and Cs134.xlsx.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 5.8 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012.

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL, dated March 7, 2012 and the following specific Sections:

71.31(a)(1) 71.31(a)(2) 71.33 71.35(a) 71.31.(c) 71.71 71.43(f) 71.51(a)(1) 71.47(a) 71.47(b) 71.47 71.31 71.51(a)(2) 71.73

3. LA-UR-13-22934, Initial MCNP 6 Release Overview - MCNP6 version 1.0, Los Alamos National Laboratory, T. Goorley, et al., April 2013.

4. ORIGEN-S Data Libraries, ORNL/TM-2005/39, Volume 3, Section M6, January 2009.

5. ORIGEN-S: SCALE System Module To Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup And Decay, And Associated Radiation Source Terms, I. C. Gauld, O. W. Herman and R. M. Westfall, ORNL/TM-2005/39, Volume 2, Section F7, January 2009.

6. RTL-001-CALC-SH-0101, Rev. 1, "Source Term Characterization for the RT-100" (PROPRIETARY)

7. RTL-001-CALC-SH-0201, Rev. 5, "Shielding Evaluation of the RT-100 Transport Cask" (PROPRIETARY)

8. CN-13039-502, "Updated Resin/Filter Shielding Evaluation of the RT-100 Transport Cask"

9. ORNL/TM-2005/39, Volume III, Section M8, Standard Composition Library, L.M. Petrie, P.B. Fox and K. Lucius, January 2009.

10. PNNL-15870, Compendium of Material Composition Data for Radiation Transport Modeling, R.G. Williams III, C.J. Gesh and R.T. Pagh, April 2006.

11. Faujasite-Na Mineral Data, Retrieved August 27, 2013, Retrieved from http://webmineral.com/data/Faujasite-Na.shtml.

12. J. Conlin, et al., "Listing of Available ACE Data Tables, LA-UR-13-21822 Rev-2, Los Alamos National Laboratory, Dec 2013.

13. ANSI/ANS 6.1.1-1997, "American National Standard for Neutron and Gamma Flux-To-Dose Conversion Factors," American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL, www.ans.org.

14. J. H. Hubbell and S. M. Seltzer, "Tables of X---Ray Mass Attenuation Coefficients and Mass-

--Energy Absorption Coefficients from 1keV to 20MeV for Elements Z = 1 to 92 and 48 Additional Substances of Dosimetric Interest," Radiation Physics Division, PML, NIST. Jul 2004. Web Nov 2014. http://www.nist.gov/pml/data/xraycoef/index.cfm Robatel Technologies, LLC Page 5-106

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

15. CN-21004-501, Rev. 1, RT-100 Supplemental Filter Shielding Evaluation (PROPRIETARY)

16. CN-21004-502, Rev. 0, RT-100 Irradiated Hardware Shielding Evaluation (PROPRIETARY)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

6. CRITICALITY EVALUATION This Section is NOT APPLICABLE. The RT-100 is not designed to transport fissile material subject to the requirements of 10 CFR Part 71 Sections 71.55 or 71.59. Therefore, no criticality evaluation is necessary for the SAR of the RT-100.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

7. PACKAGE OPERATIONS Chapter 6 describes the RT-100 operations during loading and preparation for shipment. These operations delineate the fundamental steps needed to ensure that the RT-100 is properly prepared for transport, and ensure that the operations are consistent with the previous sections of this application.

RT-100 loading and preparation operations are consistent with maintaining occupational radiation exposures as low as reasonably achievable (ALARA), as required by the Standards for Protection Against Radiation in 10 CFR 20.1101(b) [Ref. 8]. RT verifies that the operating controls and procedures meet the requirements of 10 CFR Part 71; furthermore, the operating procedures are adequate to ensure the RT-100 is operated in a manner consistent with the procedures and requirements of this Safety Analysis Report. The RT operating controls and procedures ensure the transportation safety with respect to the United States (US) requirements for transportation packages for radioactive material.

A separate operations manual is to be prepared for the RT-100 to describe the operational steps in greater detail. The regulatory requirements for the operating controls and procedures evaluation from 10 CFR Part 71 [Ref. 2] include the following issues:

o Application identifies the established codes and standards used for the operating procedures in accordance to 10 CFR Part 71 Section 71.31(c) [Ref. 2].

o Application for a fissile material is not applicable for the RT-100 since it will not be used for fissile material transport.

o RT-100 shall be transported as a Type B shipment B(U)-96 exclusive use shipment.

o The shipper shall ensure that the routine determination of 10 CFR 71.87 [Ref. 2] is met prior to each shipment. Prior to delivery of a package to a carrier, RT will send to the consignee any special instructions needed to safely open the package and use it in accordance with 10 CFR 20.1906(e) [Ref. 8] and 10 CFR Part 71 Section 71.89 [Ref. 2].

o The operating procedures meet the regulatory requirements listed in the Preparation of Empty Package for Transport Section.

o The operating procedures are adequate to ensure that the package will be operated in accordance with the Safety Analysis Report.

Input from the other sections of this application are used to develop the RT-100 operational controls and procedures. Information flow for the operating procedures evaluation is shown in Figure 7-1.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 7-1 Information Flow for the Operating Procedures Review General Structural Information Evaluation

Operational

Closure Requirements Features

Lifting Configuration

General Restrictions

Tie-Down Configuration

Tamper Indicating

Handling Restrictions Device

Contents Thermal Containment Shielding Criticality Evaluation Evaluation Evaluation Evaluation

Temperatures

Closure

Dose Rates

Not Applicable

Pressures Requirements

Streaming Paths

Assembly

Pre-Shipment Tests Verification

Leakage Rate Operating Procedures Review Package Loading Package Unloading General Requirements

Preparation for Loading

Receipt of Package

Contamination Limits

Loading

Contents Removal

Verification of Contents Removal

Preparation for Transport Acceptance Tests and Maintenance

Periodic Testing

Replacement Component Testing Robatel Technologies, LLC Page 7-2

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.1 Package Loading Section 7.1 describes loading-related preparations, tests, and inspections of the package. These actions include the inspections made before loading the package to ensure the package is not damaged, and that radiation and surface contamination levels are within allowable limits.

The procedures for loading of the RT-100 are defined as follows:

7.1.1 Preparation for Loading 7.1.2 Loading of the RT-100 7.1.3 Preparation for Transport 7.1.1 Preparation for Loading The following actions are taken prior to loading operations:

1. RT-100 is surveyed for surface contamination to ensure it is within allowable limits. If the package exceeds contamination limits, the RT-100 must be decontaminated prior to the next step.

2. A visual inspection is performed to determine if any component damage has occurred that would prevent safe performance of the package under NCT and HAC. Any damaged/out-of-specification components are repaired or replaced.

3. A pre-loading briefing is conducted with the loading team with the purpose of reviewing all procedures, and ensuring everyone is cognizant of the loading requirements and the associated safety measures. Pre-loading actions include appropriate ALARA measures and contamination control.

4. The package contents data is reviewed to ensure the contents meet the Certificate of Compliance.

5. The following conditions must be met for safe handling of the RT-100:

o All operating instructions/procedures outlined in the Safety Analysis Report must be followed.

o RT-100 shall not be lifted via the lifting rings on the upper impact limiter.

o RT-100 shall only be lifted in the vertical position.

o RT-100 shall not be placed in an upside down position at any time.

o RT-100 shall not be handled while tied down.

Loading of the RT-100 can take place on or off the trailer, and with or without the lower impact limiter attached. Additionally, loading can take place via the primary or secondary lid. Figure 7.1.1-1 illustrates the process flow for these steps. Subsequent sections describe these steps in greater detail. If the RT-100 remains on the trailer, the trailer and lower impact limiter must be protected against contamination during the procedure.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.1.1.1 7.1.1.3 Upper Impact Removal of Quick-Disconnect Limiter Removal Valve Cover Plate Loading can 7.1.1.2 take place via Optional Primary or Secondary Lid.

Loading Steps Does the cask need to be removed from NO the trailer for loading? 7.1.1.4 YES Removal of the Is it necessary to remove the Lower Primary Lid Impact Limiter for loading?

NO 7.1.1.5 YES Removal of the Secondary Lid 7.4.1 Lower Impact Limiter Removal 7.4.2 Package Removal from Trailer Figure 7.1.1-1 Preparation for Loading Process Flowchart Robatel Technologies, LLC Page 7-4

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.1.1.1 Upper Impact Limiter Removal Remove the upper impact limiter by following these steps:

1. Remove tamper indicating seal from the upper impact limiter aligning pin.

2. Remove the cotter pin from each upper impact limiter bolt.

3. Using appropriate tools loosen and remove all head nuts (12 Hex Head Nuts, M36) which secure the impact limiter to the cask body.

4. Install three lifting rings to the upper impact limiter. Remove the impact limiter using an appropriate lifting device attached to the three (3) lifting rings.

Note: To prevent damage, the impact limiter shall be handled and stored with care. It must be placed on a clean, flat surface in a position so that the studs and aligning pins do not touch the ground.

7.1.1.2 Optional Loading Steps o Lower Impact Limiter Removal: Refer to the actions described in Section 7.4.1.

o Package Removal from Trailer: Refer to the actions described in Section 7.4.2.

7.1.1.3 Removal of Quick-Disconnect Valve Cover Plate CAUTION: In the event of failure of the quick disconnect valve, radioactive material may be released when opening the vent port cover plate. Use caution to consider potential release of material consistent with the form of the cask contents.

Before opening the primary or secondary lid, internal and external pressure must be balanced to ensure safety, contamination control and to easily remove the lid(s). This is accomplished following these steps:

1. With appropriate tools, loosen and remove all bolts (6 Socket Head Cap Screws, M10x30) which secure the quick-disconnect valve cover plate to the primary lid.

2. Manually install and hand tighten two (2) of the six (6) bolts previously removed from the quick-disconnect valve cover plate into two (2) threaded holes specially designed for this purpose. Remove the cover plate using the bolts.

3. Vent the cask cavity by connecting the quick-disconnect valve to a leak tight ventilation system.

NOTES:

o Treat the quick-disconnect valve cover plate, its cavity surfaces, bolts and O-rings as potentially contaminated.

o The quick-disconnect valve cover plate must be set down with caution to prevent damage.

o Any defective bolts or O-rings, or those showing signs of deterioration shall be replaced with components meeting the specifications in the RT100 NM 1000-F Bill of Material (Chapter 1, Appendix 1.4, Attachment 1.4-1).

o Maintenance leakage rate testing shall be performed in accordance with Section 8.2.2.1 Robatel Technologies, LLC Page 7-5

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 prior to returning a package to service following maintenance, repair (Such as a weld repair), or replacement of components of a containment boundary.

7.1.1.4 Removal of the Primary Lid Remove the primary lid by following these steps:

1. Remove all the bolts (32 Hex Head Cap Screws, M48x170) securing the lid to the cask body.

2. Insert the three (3) lid lifting rings in the threaded holes designed for this purpose on the upper side of the primary lid.

3. Remove primary lid using suitable lifting equipment attached to the three lifting rings.

4. Inspect all bolt threads, hole threads, and O-rings for damage.

NOTES:

o Treat the primary lid, cask cavity surfaces, bolts, and O-rings as potentially contaminated.

o The primary lid shall be handled and stored with care in order to prevent damage.

o Any defective bolt or O-ring, or those showing signs of deterioration shall be replaced with components meeting the specifications in the RT100 NM 1000-F Bill of Material (Chapter 1, Appendix 1.4, Attachment 1.4-1).

o Maintenance leakage rate testing shall be performed in accordance with Section 8.2.2.1 prior to returing a package to service following maintenance, repair (Such as a weld repair),

or replacement of components of a containment boundary.

o Removal of the secondary lid is not necessary to remove the primary lid.

7.1.1.5 Removal of the Secondary Lid Remove the secondary lid by following these steps:

1. Remove all the bolts (18 Hex Head Cap Screws, M36x120) securing the secondary lid to the primary lid.

2. Insert the three (3) lid lifting rings in the threaded holes designed for this purpose in the upper side of the secondary lid.

3. Remove secondary lid using suitable lifting equipment attached to the three lifting rings.

4. Inspect all bolt threads, hole threads, and O-rings for damage.

NOTES:

o Treat the secondary lid, cask cavity surfaces, bolts, and O-rings as potentially contaminated.

o The secondary lid shall be handled and stored with care in order to prevent damage.

o Any defective bolt or O-ring, or those showing signs of deterioration shall be replaced with components meeting the specifications in the Bill of Material (Appendix 1.4).

o Maintenance leakage rate testing shall be performed in accordance with Section 8.2.2.1 prior to returing a package to service following maintenance, repair (Such as a weld repair),

or replacement of components of a containment boundary.

7.1.2 Loading of the RT-100 Robatel Technologies, LLC Page 7-6

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Section 7.1.2 addresses the actions for loading of the RT-100. Figure 7.1.2-1 describes the process flow for these steps. Subsequent sections describe these steps in detail.

Prior to loading, shipper ensures the 7.1.2.1 package contents meet the requirements Content Loading of the RT-100:

1) Certificate of Compliance Replacement of

2) Loading Table applicable lid.

(Section 7.6.1.1)

3) Hydrogen Generation Evaluation (Section 7.5) 7.1.2.2 7.1.2.3 Primary Lid Secondary Lid Replacement Replacement 7.1.2.4 Quick-Disconnect Valve Cover Plate Replacement Figure 7.1.2-1 Loading of the RT-100 Process Flowchart 7.1.2.1 Content Loading Load the cask by following these steps:

1. Prior to loading of the RT-100, the following conditions shall be met:

a. Package contents meet the requirements of the RT-100 Certificate of Compliance

b. Package contents meet the requirements of Appendix 7.6 and its loading table addressed in Section 7.6.1.2

c. Package contents meet the requirements of the hydrogen generation evaluation described in Section 7.5

2. Ensure the contents, secondary container, and packaging are chemically compatible (i.e.,

will not react to produce flammable gases).

3. Inspect RT-100 interior for any damage, loose material or moisture.

4. Radioactively contaminated liquids may be pumped out or removed using absorbent material.

5. Removal of any material from inside the cask is performed under the supervision of qualified health physics personnel, and in accordance with health & safety requirements.

6. Pre-position any shoring necessary to shore/brace the liner during normal transit.

7. Inspect the applicable sealing surface before loading. Clean if necessary.

Note: Cleanliness of the sealing surface will have a direct effect on leak testing results.

8. Place disposable liner, drums, or other containers into the pre-positioned shoring; install additional shoring or bracing as necessary to restrict movement of contents during normal Robatel Technologies, LLC Page 7-7

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 transport.

9. Process liner as necessary and cap the liner as required.

7.1.2.2 Primary Lid Replacement The primary lid is attached by following these steps:

1. Clean and inspect the O-rings. Lubricate if necessary. Contact RT if any damage, crack, or condition is noted that can prevent O-rings from sealing properly.

Note: This step may be performed prior to cask loading.

2. If not already present, install the three (3) lifting rings on the primary lid in the threaded holes designed for this purpose.

3. Place the primary lid on the cask. The lid is positioned by an aligning pin to ensure the proper placement of the primary lid on the cask body.

4. Lubricate (if necessary) the thirty-two (32) bolt and hole threads.

5. Install the bolts with washers.

6. Tighten the bolts using the star pattern method to ensure evenly distributed pressure on the primary lid and cask body.

a. Use an initial torque of 400 N-m +/- 10%.

b. Use a final torque of 850 N-m +/- 10%.

7. Remove the three (3) lifting rings from the primary lid.

8. Refer to Section 8.2.2.2 or Section 8.2.2.3 for instructions regarding Pre-shipment Leak Testing.

7.1.2.3 Secondary Lid Replacement The secondary lid is attached by following these steps:

1. Clean and inspect the O-rings. Lubricate if necessary. Contact RT if any damage, crack, or condition is noted that can prevent O-rings from sealing properly.

Note: This step may be performed prior to cask loading.

2. If not already present, install the three (3) lifting rings on the secondary lid in the threaded holes designed for this purpose.

3. Place the secondary lid on the primary lid. The lid is positioned by an aligning pin to ensure the proper placement of the secondary lid on the primary lid.

4. Lubricate (if necessary) the eighteen (18) bolt and hole threads.

5. Install the bolts with washers.

6. Tighten the bolts using the star pattern method to ensure evenly distributed pressure on the secondary and primary lids.

a. Use an initial torque of 150 N-m +/- 10%.

b. Use a final torque of 350 N-m +/- 10%.

7. Remove the three (3) lifting rings from the secondary lid.

8. Refer to Section 8.2.2.2 or Section 8.2.2.3 for instructions regarding Pre-shipment Leak Testing.

7.1.2.4 Quick-Disconnect Valve Cover Plate Replacement Robatel Technologies, LLC Page 7-8

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 The quick-disconnect valve cover plate is replaced in the following manner:

1. Clean and inspect the O-rings. Lubricate if necessary. Contact RT if any damage, crack, or condition is noted that can prevent O-rings from sealing properly.

Note: This step may be performed prior to cask loading.

2. Install two (2) of the previously removed cover plate bolts (Socket Head Cap Screw, M10x30) in the threaded holes on the cover plate. Manually place the cover plate on the primary lid. Subsequently, remove these two (2) bolts.

3. Lubricate (if necessary) the six (6) quick-disconnect valve cover plate bolt and hole threads.

4. Secure the quick-disconnect valve onto the primary lid using plate bolts and washers.

5. Tighten bolts using the star pattern method to ensure consistent pressure on the quick-disconnect valve and primary lid.

a. Tighten the bolts by hand to compress the O-rings (no specific torque required).

b. Use a final torque of 27 N-m +/- 10%.

6. Refer to Section 8.2.2.2 or Section 8.2.2.3 for instructions regarding Pre-Shipment Leak Testing.

7.1.3 Preparation for Transport The following general requirements are completed prior to final transport of the RT-100:

1. Contamination survey completed on the external surfaces to confirm that non-fixed (removable) radioactive contamination is as low as reasonably achievable, and is within the limits specified in 49 CFR 173.443 [Ref. 3], as required by 10 CFR 71.87 [Ref. 2]. If contamination is within limits, preparation for transport may be conducted. If contamination exceeds the limits, the RT-100 must be decontaminated until the contamination limits are met.

2. Measure the exterior gamma radiation levels following RIS 13-04 to ensure these do not exceed 200 millirem per hour (2 mSv/h) at any point on the vertical planes projected from the outer edges of the trailer, on surface of the impact limiter at the axial center line if the package, and on the lower external surface of the trailer, 10 millirem per hour (0.1 mSv/h) at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the trailer (excluding the underside of the trailer) and 2 millirem per hour (0.02 mSv/h) in the tractor cab, in accordance with 49 CFR 173.441 and 10 CFR 71.47. Measurements shall be made at the axial mid-plane of the cask and below the cask end on the lower external surface of the trailer. Also measure the neutron radiation to ensure there is not unexpected neutron sources in the content.

o Caution: Ensure calibration is current and address radiation detector uncertainty when measuring pre-shipment dose rates.

3. If necessary, install the lower impact limiter and load the package on the trailer:

o Replacement of Lower Impact Limiter: Refer to Section 7.4.3 o Reloading the Package onto the Trailer: Refer to Section 7.4.4 Robatel Technologies, LLC Page 7-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 7.1.3-1 describes the process flow for transport preparation. Subsequent sections describe these steps in greater detail.

Was the lower YES 7.4.3 impact limiter previously Replacement of Lower NO removed? Impact Limiter Was the package YES 7.4.4 previously removed from Reloading the Package NO the trailer? onto the Trailer 7.1.3.1 Replacement of the Upper Impact Limiter 7.1.3.2 Verification for Transport Figure 7.1.3-1 Preparation for Transport Process Flowchart 7.1.3.1 Replacement of Upper Impact Limiter Install the upper impact limiter using the following steps:

1. Inspect the cask body flange. Inspect each threaded stud for cleanliness, defects, and lubrication - Replace if necessary in accordance with Section 7.4.5.3.

Note: This step may be performed prior to cask loading.

2. Lift the upper impact limiter using an appropriate lifting device fixed to its three (3) lifting rings.

3. Position the upper impact limiter using the two (2) aligning pins.

4. Lower the upper impact limiter slowly and cautiously to prevent any damage to threaded studs and aligning pins.

5. Disconnect the lifting device from the upper impact limiter.

6. Disconnect or render inoperable the 3 lifting rings on the upper impact limiter.

7. Secure the cask to the upper impact limiter by installing twelve (12) head nuts on the threaded studs. The head nuts do not require any specific torque. The nuts must be in contact with the cask attachment ring.

8. Install a cotter pin on each threaded stud.

9. Install a tamper indicating seal on the upper impact limiter aligning pin.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.1.3.2 Verification for Transport The following actions are confirmed prior to shipment of a loaded package.

o Licensed consignee who expects to receive the package containing materials in excess of Type A quantities specified in 10 CFR 20.1906(a) [Ref. 8] meets and follows the requirements of 10 CFR 20.1906 [Ref. 8], as applicable.

o Before delivery of a package to a carrier for transport, the shipper shall ensure that any special instructions needed to safely open the package have been sent to, or otherwise made available to, the consignee for the consignee's use in accordance with 10 CFR 20.1906(e)

[Ref. 8].

o Trailer placarding and cask labeling meet DOT specifications (49 CFR 172).

o Provisions of 10 CFR 71.87 [Ref. 2] are met.

o Radiation dose rates are in accordance with 10 CFR 71.47 [Ref. 2]

< 200 mrem/hr at package surface.

< 10 mrem/hr at 2 meters from the vertical side of the trailer.

<2 mrem/hr at the cab of the tractor o No temperature survey is required. The thermal evaluation demonstrates that the temperature requirement of 10 CFR 71.43(g) [Ref. 2] is met.

o Security seals are properly installed as required by 10 CFR 71.43(b) [Ref. 2]

o Inspect the exterior of the cask for damage prior to shipping a loaded package. Contact RT if damage is present.

o Ensure that the RT-100 is correctly tied-down to the trailer.

7.2 Package Unloading Section 7.2 addresses the unloading operations for the RT-100.

The procedures for unloading the RT-100 are defined as follows:

7.2.1 Receipt of Package from Carrier 7.2.2 Removal of Contents 7.2.1 Receipt of Package from Carrier The following actions are taken upon receipt of the RT-100 from the carrier:

1. User shall follow the applicable requirements of 10 CFR 20.1906 [Ref. 8] Procedures for Receiving and Opening Packages when the package contains radioactive material in excess of Type A quantities. Corresponding packages are identified by review of the shipping papers.

2. Any special instructions provided by the shipper in accordance with 10 CFR 71.89 [Ref. 2]

are reviewed and followed.

3. Perform visual examination of the unopened cask. Any damage is reported to the shipper and actions taken to replace or repair any components that would jeopardize the integrity of the RT-100.

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4. Inspect the tamper indicating seal at the upper impact limiter aligning pin. Shipper is notified and the shipment may be rejected by the consignee if the tamper seal has been removed or tampered with in any way. At consignee discretion, the consignee may proceed to accept the RT-100 contents if the tamper indicating seal was damaged during shipment.

5. Conduct contamination and radiation dose rate surveys to determine if the levels are compliant with the DOT and NRC. If either of these surveys exceeds the limits, the shipper is notified immediately, and the shipper collaborates with consignee; or the appropriate regulatory authorities and DOT, to resolve the issue.

6. Measure the exterior gamma radiation levels following RIS 13-04 to ensure these do not exceed 200 millirem per hour (2 mSv/h) at any point on the vertical planes projected from the outer edges of the trailer, on surface of the impact limiter at the axial center line if the package, and on the lower external surface of the trailer, 10 millirem per hour (0.1 mSv/h) at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the trailer (excluding the underside of the trailer) and 2 millirem per hour (0.02 mSv/h) in the tractor cab, in accordance with 49 CFR 173.441 and 10 CFR 71.47. Measurements shall be made at the axial mid-plane of the cask and below the cask end on the lower external surface of the trailer. Also measure the neutron radiation to ensure there is not unexpected neutron sources in the content.

7.2.2 Removal of Contents Unloading is conducted with the RT-100 in a vertical position in an approved licensed radioactive materials facility. If desired, the RT-100 may be removed from the trailer and the lower impact limiter may be removed. Follow the instructions in Section 7.4 for these optional steps.

The following procedures are used to remove the contents from the RT-100:

1. Remove the upper impact limiter in accordance with Section 7.1.1.1.

2. Remove the quick-disconnect valve cover plate and balance the interior and exterior pressures in accordance with Section 7.1.1.3.

o Caution: In the event of failure of the quick disconnect valve, radioactive material may be released when opening the vent port cover plate. Use caution to consider potential release of material consistent with the form of the cask contents.

3. Remove applicable lid (handle as contaminated):

o Removal of the Primary Lid: Section 7.1.1.4 o Removal of the Secondary Lid: Section 7.1.1.5

4. Use appropriate equipment to remove the contents.

5. Inspect RT-100 interior for any damage, loose material, or moisture. Clean seal surfaces.

Contact RT in case of damaged O-ring.

NOTES:

o Care should be taken not to damage the cask or the sealing surfaces.

o Remove any material from inside the cask under the supervision of qualified health Robatel Technologies, LLC Page 7-12

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 physics personnel, and by following appropriate health & safety protocols.

o Radioactively contaminated liquids may be pumped or removed by absorbent material.

7.3 Preparation of Empty Package for Transport Section 7.3 describes the operations used to certify that the empty package is safe for transportation in accordance with 49 CFR 173.428, Empty Class 7 (Radioactive) Material Packaging [Ref. 4].

These operations must be completed before shipment of the empty RT-100:

1. Confirm the cavity is empty of contents as far as practicable

2. Survey the lid, quick connect cover and interior:

o Decontaminate if the limits of 49 CFR 173.428(d) [Ref. 4] are exceeded.

3. Replace and secure the applicable lid:

o Primary Lid Replacement 7.1.2.2 o Secondary Lid Replacement 7.1.2.3

4. Replace and secure the quick-disconnect valve cover in accordance with Section 7.1.2.4.

5. Decontaminate the exterior surfaces of the cask, as necessary.

6. If previously removed, replace lower impact limiter in accordance with Section 7.4.3.

7. If previously removed, replace RT-100 on the trailer in accordance with Section 7.4.4.

8. Replace upper impact limiter in accordance with Section 7.1.3.1.

9. Inspect the exterior and confirm it is undamaged and unimpaired.

7.4 Other Operations Section 7.4 describes provisions for any special operational controls.

7.4.1 Lower Impact Limiter Removal The lower impact limiter may be removed by following these steps:

1. Remove the cotter pin from each lower impact limiter threaded stud.

2. Using appropriate tools loosen and remove all the head nuts which secure the impact limiter to the cask body.

3. Remove the cask from the trailer following the sequence described in Section 7.4.2.

4. Install three lifting rings to the lower impact limiter using three (3) 120° distant studs.

5. Remove the impact limiter using an appropriate lifting device attached to the three (3) lifting rings.

Note: To prevent damage, the impact limiter shall be handled and stored with care. It must be placed on a clean, flat surface in a position so that the studs and aligning pins do not touch the ground.

7.4.2 Package Removal from Trailer The RT-100 may be removed from the trailer by following these steps:

1. Verify transport trailer is in appropriate area for unloading.

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2. Disconnect tie-down system and render the holes inoperable so that they cannot be used for lifting of the packaging.

3. If not already complete, remove upper impact limiter in accordance with Section 7.1.1.1.

4. Inspect condition of the cask lifting pocket welds.

5. Install Lifting Yoke. (RT-100 Lifting Yoke must be handled with its two (2) suitable lifting rings connected to the lifting crane.)

a. Engage two (2) lifting yoke arms in the cask lifting pockets (Figure 7.4.2-1).

b. Connect arms in the lifting pockets by inserting the pin in each arm (Figure 7.4.2-2).

c. Secure lifting yoke arms by inserting lock pin in the pin hole below the cask lifting pockets, and secure the lock pin with a cotter pin (Figure 7.4.2-3).

6. With or without the lower impact limiter attached, lift the RT-100 cask from the transport trailer (Figure 7.4.2-4).

7. Place the RT-100 cask (and lower impact limiter) in an approved storage area. Care should be taken to prevent the cask assembly and the lifting equipment from any damage.

8. Disconnect the lifting yoke from the cask.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 7.4.2-1 Lifting Yoke Arm Figure 7.4.2-2 Lifting Yoke Positioned on Cask Connections Figure 7.4.2-3 Lifting Yoke Secured with Locking Pin Figure 7.4.2-4 Assembled Cask Ready to Lift (shown with Lower Impact Limiter installed)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.4.3 Replacement of Lower Impact Limiter Replace the lower impact limiter by following these steps:

1. Verify that the lower impact limiter is correctly positioned and located in an appropriate area for installing the cask. The installation of the lower impact limiter can be conducted with the limiter already positioned on the transport trailer.

2. Inspect and clean the lower impact limiter surface.

3. Inspect each threaded stud for cleanliness, defects and lubrication - replace if necessary in accordance with Section 7.4.5.3.

4. Install the lifting yoke on the cask as written in Section 7.4.2 (upper impact limiter shall not be installed on the cask).

5. Lift cask with appropriate lifting equipment.

6. Inspect and clean (if necessary) bottom surface of cask.

7. Position the cask body using the two (2) lower impact limiter aligning pins.

8. Using caution, slowly lower the cask body to prevent any damage to the lower impact limiter and the cask body.

9. Secure cask on the lower impact limiter by installing the twelve (12) nuts on the threaded studs. The head nuts do not require any specific torque. The nuts must be in contact with the cask attachment ring.

10. Install a cotter pin on each of the 12 threaded studs.

7.4.4 Reloading the Package onto the Trailer The RT-100 may be reloaded onto the trailer by following these steps:

1. Verify the transport trailer is located in an appropriate area for loading. Figure 7.4.4-1 shows an example trailer illustration.

2. Install the lifting yoke on the cask as written in Section 7.4.2 (upper impact limiter shall not be installed on the cask).

3. Lift the RT-100 cask onto the transport trailer as shown in Figure 7.4.4-2. Position the tie-down arms along the trailer as shown in Figure 7.4.4-2.

4. Disconnect the lifting yoke from the cask as described in Section 7.4.2.

5. The lifting pockets must be rendered inoperable prior to transport. Acceptable options for this step include: inserting the lifting pins in the lifting pockets and securing with the lock pins, or inserting and securing dedicated pins into the lifting pockets.

6. Install the upper impact limiter as described in Section 7.1.3.1.

7. Connect the four (4) tie-down arms to the trailer by the tie-down equipment.

8. Tighten the tie-down equipment to ensure no movement of the RT-100 in any direction.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 7.4.4-1 Example Trailer Illustration Figure 7.4.4-2 Loading of the RT-100 on Transportation Trailer Robatel Technologies, LLC Page 7-17

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.4.5 Tightening of Components Section 7.4.5 addresses issues associated with components that require tightening.

7.4.5.1 Tightening Torques Table 7.4.5-1 provides descriptive information and torque values for lid bolts. Table 7.4.5-2 provides descriptive information and torque values for other parts.

Table 7.4.5-1 Lid Bolt Tightening Torques Description Dimensions Qty. Tightening Tolerance Torque [N-m]

Primary lid bolts HHCS M48x170 32 850 +/- 10%

Secondary lid bolts HHCS M36x120 18 350 +/- 10%

Quick disconnect valve disconnect valve cover plate SHCS M10x30 6 27 +/- 10%

bolts Table 7.4.5-2 Tightening Torques - Other Parts Description Dimensions Qty. Tightening Tolerance Torque [N-m]

Primary Lid Aligning Pin M42 1 850 +/- 10%

Secondary Lid Aligning Pin M24 1 200 +/- 10%

Impact Limiter Studs M36 24 N/A¹ -

Impact Limiter Nuts HM36 24 N/A¹ -

Quick Disconnect Valve M10 1 10 +/- 10%

Cover Plate Leak Test Port Plug Quick Disconnect Valve G 3/8 1 50 +/- 10%

Primary and Secondary Lid M20 2 100 +/- 10%

Leak Test Port Plug

¹ The lower and upper impact limiter nuts do not require any specific torque to ensure function. They are hand-torqued when assembled. Thus, no control of the tightening torque on these nuts is necessary.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.4.5.2 Threaded Bolts - Tightening Methods and Equipment Threaded bolts and parts must be tightened to the torque limits specified in Table 7.4.5-1 and Table 7.4.5-2, using a wrench that displays the torque applied during tightening.

1. Bolts and bolt holes are cleaned prior to use.

2. If necessary, bolts and nuts are lubricated before tightening.

3. Bolts on the primary and secondary lids are tightened via the star pattern in a minimum of two stages:

a. Use an initial torque of approximately 50% of required torque.

b. Use a final torque of the full required torque.

7.4.5.3 Replacement of the Impact Limiter Threaded Studs The impact limiter threaded studs may be replaced in the following manner:

1. Remove the spring pin securing the threaded stud.

2. Unscrew the threaded stud and remove it.

3. Install a new threaded stud and tighten it by hand to align the stud and the embossing holes (no specific torque required).

4. Install a new spring pin to secure the threaded stud.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.5 Hydrogen Buildup in RT-100 Transport Cask The RT-100 is designed for a maximum decay heat of 200 Watts, as described in Section 1.2.2.8.

The rate of hydrogen gas generation must also be considered when evaluating the heat load. The method for calculating the hydrogen gas generation is described in Section 4.4. Two evaluation methods are described in the following sections - a simplified model (Section 7.5.1) which is used for most shipments but is limited to certain restrictions, and an analytical model (Section 7.5.2) used for other cases. An example calculation using the analytical model is detailed in Section 7.5.3.

Package material and content that can generate flammable gas shall be appropriately assigned as part of the ionic resin bead waste or polyethylene container when using the Loading Curve (Figure 7.5-1 and Table 7.5.1-1) or detailed analysis (Section 7.5.2) to determine acceptable hydrogen gas generation-related parameters of shipping time and decay heat. For example, waste filters (made of material other than polypropylene or polyethylene) shall be grouped as ionic bead waste and wood shoring would be grouped as part of the polyethylene container. If filters are made of polyethylene or polypropylene, they are to be included in the secondary container volume for the hydrogen gas generation detailed analysis. Activated hardware materials are not hydrogen generating and do not retain water that cannot be evacuated like filters and resin wastes. As a result, the only effect hardware materials have on the hydrogen buildup calculations in the RT-100 is if they are mixed with resin & filter contents. In this scenario, the volume occupied by the hardware contents in the cavity must be accounted for. If no hydrogenous materials are included in the contents, including the liner and shoring, the hydrogen gas buildup calculations are not required.

7.5.1 Hydrogen Gas Generation - Simplified Model used to develop Loading Curve Using the equations derived in Chapter 4, Section 4.4, the decay heat limit versus waste volume can be determined for a limit of 5% in the cavity free volume. Figure 7.5.1-1 provides a curve illustrating the waste volume to decay heat value that would result in the generation of a flammable gas mixture within 10 days assuming that all decay heat is absorbed by the waste material and the polyethylene container. The calculation assumes that the hydrogen generation occurs over a period of time that is twice the allowable shipping time. For most shipments, this simplified graphical model (Loading Curve) can be used to determine the maximum heat load.

Use the following procedure to confirm the decay heat of the cask contents meet the requirements of NUREG/CR-6673 [Ref. 12]:

1. Confirm all the restrictions listed in Table 7.5.1-1 are met. Confirm the secondary container is listed in Table 7.5.1-2, or is a container of equivalent material volume.

Confirm the shoring volume does not exceed the allowable volume. If these restrictions cannot be met, the analytical model described in Section 7.5.2 must be used.

2. Identify the waste volume in cubic feet. Waste volume in the simplified model include Robatel Technologies, LLC Page 7-20

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 resins, filters, and activated hardware (low-density and high-density). Note that for this simplified model, activated hardware waste in the contents is grouped as ionic bead waste and is assumed to be hydrogen generating.

3. Use the Loading Curve shown in Figure 7.5.1-1 to find the maximum allowable decay heat (DH,max) in Watts.

4. Confirm that the cask contents have a decay heat (DH) that is less than the maximum allowable decay heat (DH,max).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 7.5.1-1 Package Loading Curve for Hydrogen Generation - Decay Heat Limit Versus Waste Volume Table 7.5.1-1 Conditions for using Package Loading Curve (Excerpt from Table 4.4.4-1)

Condition for Shipper to use Loading Curve Waste consisting of resins/filters or mixture of resins/filters with activated hardware from 1

commercial power plants 2 Waste has been dewatered or grossly dewatered 3 No limit on moisture content of resin 4 Use of a liner (or equivalent) listed in Table 7.5.1-2 with maximum shoring volume as specified Shipment time not greater than 10 days for resin/filter shipments or mixture of resin/filter with 5

activated hardware shipments 6 Loading at temperature not exceeding 38 °C and standard pressure (1 atm) 7 Secondary containers are passively vented within the cask cavity during shipment.

8 Filters in the waste are not made of polyethylene or polypropylene.

9 Waste Volume not greater than 130 ft3.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.5.1-2 Secondary Container and Allowable Shoring Volumes (Excerpt from Table 4.4.3-2)

Volume Occupied by Allowable Shoring Volume Secondary Container Container (ft3) (cm3) (ft3) (cm3)

PL 6-80 MT 9.28 2.63E+05 20.82 5.90E+05 PL 6-80 MTIF 9.74 2.76E+05 20.36 5.76E+05 PL 6-80 FR 10.21 2.89E+05 19.89 5.63E+05 PL 6-80 FP/FEDX 11.60 3.28E+05 18.50 5.24E+05 PL 8-120 MT 11.14 3.15E+05 18.96 5.37E+05 PL 8-120 MTIF 11.60 3.28E+05 18.50 5.24E+05 PL 8-120 FR 12.06 3.42E+05 18.04 5.11E+05 PL 8-120 FP/FEDX 13.46 3.81E+05 16.64 4.71E+05 PL 8-120 CMT 13.36 3.78E+05 16.74 4.74E+05 PL 14-150 14.85 4.20E+05 15.25 4.32E+05 PL 10-160 MT 12.99 3.68E+05 17.11 4.84E+05 PL 10-160 MTIF 13.64 3.86E+05 16.46 4.66E+05 PL 10-160 FR 13.92 3.94E+05 16.18 4.58E+05 PL 10-160 FP/FEDX 15.31 4.34E+05 14.79 4.19E+05 NUHIC-55 2.78 7.88E+04 27.32 7.74E+05 NUHIC-136 11.14 3.15E+05 18.96 5.37E+05 Radlok 500 12.62 3.57E+05 17.48 4.95E+05 EL-50 16.87 4.78E+05 13.23 3.75E+05 EL-142 27.02 7.65E+05 3.08 8.72E+04 L 6-80 MT 2.27 6.42E+04 27.83 7.88E+05 L 6-80 CMT 2.61 7.38E+04 27.49 7.79E+05 L 6-80 IN-SITU 7.94 2.25E+05 22.16 6.28E+05 L 6-80 FP 2.38 6.74E+04 27.72 7.85E+05 L 6-80 FP/FEDX 2.78 7.86E+04 27.32 7.74E+05 L 8-120 MT 2.72 7.70E+04 27.38 7.75E+05 L 8-120 CMT 3.06 8.67E+04 27.04 7.66E+05 L 8-120 IN-SITU 9.52 2.70E+05 20.58 5.83E+05 L 8-120 FR 2.83 8.03E+04 27.27 7.72E+05 L 8-120 FP/FEDX 3.00 8.51E+04 27.10 7.67E+05 ES-50 0.57 1.61E+04 29.53 8.36E+05 ES-142 2.49 7.06E+04 27.61 7.82E+05 Robatel Technologies, LLC Page 7-23

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.5.2 Hydrogen Gas Generation - Analytical Model used for Detailed Analysis If Figure 7.5.1-1 is not applicable to a shipment, or if further analysis is required, the equations given in Section 4.4.5 can be used to determine the maximum allowable decay heat. Equation 4.8 and Equation 4.9 are given below for reference:

Equation 4.8 Determination of maximum shipping time based on a known decay heat:

(2.5 0 )(4.66 0.8911 )(0.8911 + )

=

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

Equation 4.9 Determination of the maximum decay heat based on a known shipment time:

(2.5 0 )(4.66 0.8911 )(0.8911 + )

, =

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

where: tmax = maximum allowable shipping time for a given decay heat to ensure the hydrogen generated during shipment does not exceed 5% [s]

DH,max = maximum allowable decay heat for a given shipping time to ensure the hydrogen generated during shipment does not exceed 5% [eV/s]

AN = Avogadros constant [6.022x1023 molecules/gmol]

Rg = gas law constant [82.05 cm3atm/gmolK]

P0 = pressure when the container is sealed [atm]

T0 = temperature when the container is sealed [K]

t = shipment time [s]

DH = decay heat of cask contents [eV/s]

VC = volume occupied by the secondary container, shoring, and polyethylene or polypropylene filters in the waste [cm3]

VWASTE = volume occupied by the ionic resin and stainless steel filters in the waste material [cm3]

VH = volume occupied by activated hardware in the waste material (combined volumes of low-density and high-density hardware) [cm3]

GTi = total radiolytic G value for the ionic resin and stainless steel filters

[molecules/100eV]

GTC = total radiolytic G value for the secondary container, shoring, and polyethylene or polypropylene filters in the waste [molecules/100eV]

GTW = total radiolytic G value for water in waste [molecules/100eV]

i = fraction of GTi that is equivalent to GFGi, flammable gas released, for the ionic resin and stainless steel filters C = fraction of GTC that is equivalent to GFGC, flammable gas released, for the secondary container, shoring, and polyethylene or polypropylene filters in the waste W = fraction of GTW that is equivalent to GFGW, flammable gas released, for water in the waste Robatel Technologies, LLC Page 7-24

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Note 1: Use of Equation 4.8 and Equation 4.9 are valid only when the conditions listed in Table 7.5.2-1 are met. Shipments are allowed only if the conditions in Table 7.5.2-1 are met.

Table 7.5.2-1 Conditions for Shipper to use the Detailed Analysis (From Table 4.4.5-1) 1 Waste consists of resins/filters or mixture of resins/filters with activated hardware from commercial power plants.

2 Waste has been grossly dewatered.

3 Secondary containers are passively vented within the cask cavity during shipment.

Use the following procedure to confirm the decay heat of the cask contents meet the requirements of NUREG/CR-6673 [Ref. 12]:

1. Determine the values of the variables P0, T0, VC, and VWASTE, and VH. The variable VH is only required if activated hardware contents are included with the waste. Initial pressure (P0) and initial temperature (T0) may be measured by the user at the time of loading. The volume occupied by the secondary container, shoring, and polyethylene or polypropylene filters in the waste (VC) and the volume occupied by the ionic resin and stainless steel filters in the waste material (VWASTE) are known.

2. Determine the values of the variables GTi, GTC, GTW, i, C, and W. G-values (GTi, GTC, GTW) and fractions (i, C, W) must be justified by the user based on waste characterization. These variables must be adjusted for the transport temperature of 80 °C, as described in Section 4.4.1.3, in order to meet the requirements of NUREG/CR-6673

[Ref. 12]. The values must also be adjusted for the appropriate alpha/gamma radiation distribution. One example of this adjustment is provided in Table 7.5.2-2 for the same G-values in the bounding case loading curve for the 80% gamma/20% alpha decay heat distribution.

3. Use one of the following methods (a or b).

a. Take the decay heat of the cask contents (DH) and solve Equation 4.8 for the maximum allowable shipping time (tmax). Confirm the shipment time (t) will be less than the maximum allowable shipping time (tmax).

b. Take the shipment time (t) and solve Equation 4.9 for the maximum allowable decay heat (DH,max). Confirm the actual decay heat of the contents (DH) is less than the maximum allowable decay heat (DH,max).

Note 2: Alternatively, the user can follow the NUREG/CR-6673 requirements to determine the shipping time to reach a hydrogen concentration of 5%. The shipping time has to be defined as 1/2 the time to reach the 5% hydrogen concentration per the requirement in NUREG/CR-6673 [Ref.

12].

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.5.2-2 G-values and -Fractions for a Range of Alpha/Gamma Decay Heat Distributions (Excerpt from Table 4.4.5-2)

Gamma Alpha Frac Material G (net gas), GT Frac Polyethylene GTC 5.06 C 1.00 0.0 1.0 Resin GTi 3.65 i 0.81 Water GTW 1.60 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.1 0.9 Resin GTi 3.40 i 0.82 Water GTW 1.49 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.2 0.8 Resin GTi 3.14 i 0.82 Water GTW 1.37 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.3 0.7 Resin GTi 2.88 i 0.83 Water GTW 1.26 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.4 0.6 Resin GTi 2.62 i 0.84 Water GTW 1.14 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.5 0.5 Resin GTi 2.37 i 0.85 Water GTW 1.03 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.6 0.4 Resin GTi 2.11 i 0.87 Water GTW 0.91 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.7 0.3 Resin GTi 1.85 i 0.89 Water GTW 0.80 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.8 0.2 Resin GTi 1.59 i 0.91 Water GTW 0.68 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.9 0.1 Resin GTi 1.34 i 0.95 Water GTW 0.57 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 0.95 0.05 Resin GTi 1.21 i 0.97 Water GTW 0.51 W 1.00(16)

Polyethylene GTC 5.06 C 1.00 1.0 0.0 Resin GTi 1.08 i 1.00 Water GTW 0.45 W 1.00(16) 16 For water, the value is set to 1.0.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.5.3 Hydrogen Gas Generation - Analytical Model Examples Example 1 calculation using the analytical model developed in Section 7.5.2 is shown below. The following two constants are known:

AN = 6.022E23 molecules/gmol Rg = 82.05 cm3atm/gmolK In this example, the user has to input the following parameters:

P0 = 1 atm T0 = 89 °F = 305 K t = 8 days = 691200 s VC = 30.1 ft3 = 8.52E05 cm3 (secondary container, shoring, and PE/PP volumes)

VWASTE = 70.0 ft3 = 1.98E06 cm3 (ionic resin and steel filter volumes)

VH = 0 ft3 = 0 cm3 (no hardware contents)

These variables are substituted into the maximum allowable decay heat equation as shown:

(2.5 0 )(4.66 0.8911 )(0.8911 + )

, =

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

[(2.5)(6.02223)(1)] x [4.66 8.525 (0.8911)(1.986) 0] x [(0.8911)(1.986) + 8.525]

=

[(82.05)(305)(691200)][0.6336(1.986) ( 0.05) + (8.52E5) ( 0.05) + 0.2575(1.98E6) ( 0.05)]

[1.5124] x [1.986] x [2.626]

=

[1.7310][(1.256) ( 0.05) + (8.525) ( 0.05) + (5.105) ( 0.05)]

Additionally, the user knows the decay heat distribution of the waste stream is 95% gamma. Using the information provided in Table 7.5.2-2, the user inputs the following G-values and fractions.

(Note that the values listed in Table 7.5.2-2 have already been adjusted for bounding NCT temperature.)

GTi = 1.21 molecules/100eV GTC = 5.06 molecules/100eV GTW = 0.51 molecules/100eV i = 0.97 C = 1.00 W = 1.00 With further substitution, the maximum allowable decay heat (DH,max), may be calculated:

[1.5124] x [1.986] x [2.626]

=

[1.7310][(1.256)(1.21)(0.97 0.05) + (8.525)(5.06)(1 0.05) + (5.105)(0.51)(1 0.05)]

[7.8136]

=

[1.73E10][(1.406) + (4.10E6) + (2.47E5)]

, = 7.87 x 1019 = 12.61 Robatel Technologies, LLC Page 7-27

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Example 2 calculation using the analytical model developed in Section 7.5.2 is shown below with activated hardware contents. The following two constants are known:

AN = 6.022E23 molecules/gmol Rg = 82.05 cm3atm/gmolK In this example, the user has to input the following parameters:

P0 = 1 atm T0 = 89 °F = 305 K t = 8 days = 691200 s VC = 30.1 ft3 = 8.52E05 cm3 (secondary container, shoring, and PE/PP volumes)

VWASTE = 70.0 ft3 = 1.98E06 cm3 (ionic resin and steel filter volumes)

VH = 30.0 ft3 = 8.52E05 cm3 (hardware volume)

These variables are substituted into the maximum allowable decay heat equation as shown:

(2.5 0 )(4.66 0.8911 )(0.8911 + )

, =

( 0 )[0.6336 ( 0.05) + ( 0.05) + 0.2575 ( 0.05)]

[(2.5)(6.02223)(1)] x [4.66 8.525 (0.8911)(1.986) 8.525] x [(0.8911)(1.986) + 8.525]

=

[(82.05)(305)(691200)][0.6336(1.986) ( 0.05) + (8.52E5) ( 0.05) + 0.2575(1.98E6) ( 0.05)]

[1.5124] x [1.136] x [2.626]

=

[1.7310][(1.256) ( 0.05) + (8.525) ( 0.05) + (5.105) ( 0.05)]

Additionally, the user knows the decay heat distribution of the waste stream is 95% gamma. Using the information provided in Table 7.5.2-2, the user inputs the following G-values and fractions.

(Note that the values listed in Table 7.5.2-2 have already been adjusted for bounding NCT temperature.)

GTi = 1.21 molecules/100eV GTC = 5.06 molecules/100eV GTW = 0.51 molecules/100eV i = 0.97 C = 1.00 W = 1.00 With further substitution, the maximum allowable decay heat (DH,max), may be calculated:

[1.5124] x [1.136] x [2.626]

=

[1.7310][(1.256)(1.21)(0.97 0.05) + (8.525)(5.06)(1 0.05) + (5.105)(0.51)(1 0.05)]

[4.4636]

=

[1.73E10][(1.406) + (4.10E6) + (2.47E5)]

, = 4.49 x 1019 = 7.19 Robatel Technologies, LLC Page 7-28

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6 Appendix The following appendices are included for Chapter 6 instruction and information. Additional steps and conditions of use for the RT-100 are as follows:

1. The maximum bulk filter/resin content density is 1.0 g/cm3 and/or the activated hardware content density shall be determined to be within the low-density hardware (2.0 g/cm3

< 7.5 g/cm3) or high-density hardware (7.5 g/cm3 9.0 g/cm3) range. The weight of free water must be excluded in this determination. The source strength density must be ensured at any point of the content. Average density by dividing the total activity by total weight is not acceptable.

2. No neutron emitting nuclides, except in trace amounts.

3. The weight of water must be excluded when determining the Ci/gram of content limit.

4. The source concentration must not exceed the Curies per gram limit determined using the method and the loading tables as prescribed in Appendix 7.6 of this Chapter.

5. The user/shipper must analyze the constituent radioactive nuclides of the content on a per-gram basis.

6. The user/shipper must determine the allowable content based on the loading tables provided in Section 7.6.1.1.

7. The user/shipper must ensure the per gram activities at any point within the content does not exceed the limit that is specified according to the loading table.

8. For resins and filters, the allowable content must be determined based on dry resin or filter media.

9. For resins and filters, the radioactive content is not to exceed 1.0 g/cm3 and the nuclear physical characteristics, i.e., the gamma attenuation coefficient of the content must not be smaller than that of the carbon material resin.

10. A comprehensive dose rate measurement is performed prior to transport of the package as described in Section 7.1.3.

11. A comprehensive dose rate measurement is performed after arrival at the destination as described in Section 7.2.1.

12. Compare pre-shipment dose rate measurements to dose rate measurements at the arrival of the package. Stop any further shipment if the measured dose rates show significant differences from the pre-shipment measurement values.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6.1 RT-100 Loading Table 7.6.1.1 RT-100 Loading Table Description The dose rate compliance of the RT-100 cask is ensured by: first, determining the percent of each radionuclide activity relative to its maximum allowable value, and then summing up this dose rate percentage for all radionuclides and assuring that the sum does not exceed 100%. In addition, the activity for the package shall be less than 3000 A2, the total decay heat output shall be less than 200 watts, and the actual content activity concentration of neutron emitters shall be less than 3.5 milliCuries/kg. This compliance is facilitated by employing a loading table that is completed by the shipper of the RT-100. An individual loading table is available for each of the three content types (filters & resins, low-density hardware, and high-density hardware), each of which has ten columns as follows:

o Actual Content Nuclide - Radionuclide entered by user into the loading table. Qualification of the content loaded into the RT-100 cask is under the responsibility of the shipper/user.

o Maximum Allowable Activity Concentration - The maximum allowed activity for the isotope being entered into the RT-100 transport cask, in mCi/kg, that is based on the methodology described in Sections 5.4, 5.5, and 5.6 and presented in either Table 7.6.1-18 or Table 7.6.1-19. The 8 explicitly analyzed isotopes in Sections 5.4, 5.5, and 5.6 are in Table 7.6.1-18 and the generic energy line isotopes are in Table 7.6.1-19.

o Actual Content Activity Concentration - The maximum activity concentration of each isotope for each content analyzed is entered by the user in units of mCi/kg. This value should be the maximum activity concentration of the content being analyzed (e.g., for mixed filters and high-density hardware shipments, the user shall input the maximum activity concentration of each isotope for the filter content on the Filters & Resins sheet, and the maximum activity concentration of each isotope for the hardware content on the High-density Hardware sheet). The user shall ensure that the bulk density of the wastes meet the bulk density requirements from 7.6.

o % of Maximum - The percent of maximum column represents the percentage of the activity concentration entered by the user for the isotope in question versus the maximum allowable activity concentration established by the methodology described in Section 5.4.4.

o A2 - The total activity equivalent to 1 A2 in Curies for the isotope entered by the user.

o Activity(i) / A2(i) - The total A2 quantity for the isotope entered by the user.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Q Value - The amount of heat energy released per unit activity, in Watts/milliCurie, for the isotope entered by the user.

o Heat Load - The total heat load contribution, in Watts, for the activity concentration and mass entered by the user for the isotope.

o Neutron Emitter? - This column indicates whether the Radionuclide entered by the user is a neutron emitter (238Pu, 239Pu, 240Pu, 242Pu, 242Cm, 243Cm, 244Cm, 248Cm, 241Am, or 252Cf).

o Actual Content Activity Concentration - For neutron emitters, the previously entered activity concentration is shown in this column.

The Maximum Allowable Activity Concentration (milliCuries/kg) column contains the maximum activity per kilogram allowed for each isotope based on the NCT and HAC dose rate limits and the most conservative response functions (mrem/hr/Curie) generated by MCNP calculations. For the eight explicitly analyzed radionuclides in Sections 5.4, 5.5, and 5.6, the Maximum Allowable Activity Concentration updates to the appropriate limit automatically based on the content mass input. The % of Maximum column is summed at the end of the column. If the sum is greater than 100%, than the inventory dose rate would potentially exceed the NCT or HAC dose rate limits, which would make the package not acceptable. If the sum is less than or equal to 100%, the package would generate an acceptable dose rate under NCT and HAC conditions. The A2 (Curie) column contains the A2 activity value in Curies for the isotope. The Activity (i) / A2(i) column is the ratio of the activity entered by the user divided by the A2 value for the isotope. The Activity (i) / A2(i) column is summed. If the value is under 3000 A2, the inventory total activity is below the containment limit and the package is acceptable under the containment limitation.

Each radionuclides activity is multiplied by the value in the Q Value (Watts/milliCurie) column and the result is automatically entered into the Heat Load (Watts) column. At the end of the column, the results are summed and compared to the 200 Watt limit. If it is below the limit, the package is acceptable under heat load limitations.

If the user-entered radionuclide is a neutron emitter, it is indicated in the Neutron Emitter?

column. For these radionuclides, the Actual Content Activity Concentration is shown in the last column. At the end of the column, the results are summed and compared to the limit of 3.5 milliCuries/kg. If the sum is below the limit, the package is considered to contain only trace amounts of neutron-emitting radionuclides.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Notes for the RT-100 Loading Table:

o Each radionuclide in the content is listed by row.

o The sum of column 4 (% of maximum) should be less than 100% for dose rate regulatory compliance.

o The sum of column 6 (C(i)/A(i)) should be less than 3000 A2 for containment regulatory compliance.

o The sum of column eight should be less than 200 Watts in order to satisfy the heat load limit.

o In the loading table, LD Hardware refers to low-density hardware and HD Hardware refers to high-density hardware.

o The Loading Summary sheet shows the status of user-entered radionuclides and activity concentrations relative to the four acceptance criteria (shielding, containment, heat load, and neutron limit).

o A basic RT-100 Loading Table Procedure is provided in Section 7.6.1.2.

o Loading examples are provided in Sections 7.6.1.3, 7.6.1.4, 7.6.1.5, and 7.6.1.6.

o Examples of the RT-100 Loading Table format are provided below in Table 7.6.1-1 (Filters and Resins Sheet), Table 7.6.1-2 (LD Hardware Sheet), Table 7.6.1-3 (HD Hardware Sheet), Table 7.6.1-4 and (Loading Summary Sheet). Note that these examples are presented with arbitrary masses, volumes, radionuclides, and activity concentrations for illustration purposes only.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-1 RT-100 Loading Table Illustration - Filter and Resin Sheet 15.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 215.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) ag110m 1.00E+03 1.00E+00 0.10 11 1.95E-02 1.67E-05 3.59E-03 N cs137 1.84E+05 1.00E+00 0.00 16 1.34E-02 5.04E-06 1.08E-03 N co58 1.49E+04 1.00E+00 0.01 27 7.96E-03 5.99E-06 1.29E-03 N co60 5.18E+02 2.00E+02 38.58 11 3.91E+00 1.54E-05 6.63E-01 N cs134 1.21E+04 1.00E+00 0.01 19 1.13E-02 1.02E-05 2.19E-03 N fe59 1.40E+03 1.00E+00 0.07 24 8.96E-03 7.74E-06 1.66E-03 N mn54 2.96E+04 1.00E+00 0.00 27 7.96E-03 4.98E-06 1.07E-03 N zn65 4.54E+03 1.00E+00 0.02 54 3.98E-03 3.50E-06 7.52E-04 N i131 1.71E+04 1.00E+00 0.01 19 1.13E-02 3.40E-06 7.31E-04 N pu239 1.14E+08 1.00E+00 0.00 0.027 7.96E+00 3.11E-05 6.68E-03 Y 1.00E+00 Table 7.6.1-2 RT-100 Loading Table Illustration - Low-density Hardware Sheet 10.00 LD Hardware Waste Volume (ft 3)

Actual Content Loading Evaluation 765.00 LD Hardware Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) ag110m 6.24E+02 1.00E+00 0.16 11 6.95E-02 1.67E-05 1.28E-02 N cs137 1.58E+05 1.00E+00 0.00 16 4.78E-02 5.04E-06 3.86E-03 N co58 9.09E+03 1.00E+00 0.01 27 2.83E-02 5.99E-06 4.58E-03 N co60 3.23E+02 5.00E+01 15.50 11 3.48E+00 1.54E-05 5.90E-01 N cs134 7.78E+03 1.00E+00 0.01 19 4.03E-02 1.02E-05 7.80E-03 N fe59 8.97E+02 1.00E+00 0.11 24 3.19E-02 7.74E-06 5.92E-03 N mn54 2.10E+04 1.00E+00 0.00 27 2.83E-02 4.98E-06 3.81E-03 N zn65 2.93E+03 1.00E+00 0.03 54 1.42E-02 3.50E-06 2.68E-03 N i131 1.71E+04 1.00E+00 0.01 19 4.03E-02 3.40E-06 2.60E-03 N pu239 1.14E+08 1.00E+00 0.00 0.027 2.83E+01 3.11E-05 2.38E-02 Y 1.00E+00 Table 7.6.1-3 RT-100 Loading Table Illustration - High-density Hardware Sheet 10.00 HD Hardware Waste Volume (ft 3)

Actual Content Loading Evaluation 2275.00 HD Hardware Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) ag110m 5.71E+02 1.00E+00 0.17 11 2.07E-01 1.67E-05 3.79E-02 N cs137 1.20E+05 1.00E+00 0.00 16 1.42E-01 5.04E-06 1.15E-02 N co58 8.57E+03 1.00E+00 0.01 27 8.43E-02 5.99E-06 1.36E-02 N co60 3.11E+02 5.00E+01 16.09 11 1.03E+01 1.54E-05 1.75E+00 N cs134 7.12E+03 1.00E+00 0.01 19 1.20E-01 1.02E-05 2.32E-02 N fe59 8.71E+02 1.00E+00 0.11 24 9.48E-02 7.74E-06 1.76E-02 N mn54 2.02E+04 1.00E+00 0.00 27 8.43E-02 4.98E-06 1.13E-02 N zn65 2.79E+03 1.00E+00 0.04 54 4.21E-02 3.50E-06 7.96E-03 N i131 1.71E+04 1.00E+00 0.01 19 1.20E-01 3.40E-06 7.73E-03 N pu239 1.14E+08 1.00E+00 0.00 0.027 8.43E+01 3.11E-05 7.07E-02 Y 1.00E+00 Robatel Technologies, LLC Page 7-33

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-4 RT-100 Loading Table Illustration - Loading Summary Sheet Robatel Technologies, LLC Page 7-34

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6.1.2 RT-100 Loading Table Procedure Depending on the type of contents, the respective loading table sheet shall be filled out for: Filters

& Resins, LD Hardware, or HD Hardware using the following steps. Unused loading table sheets shall remain blank, if only HD Hardware is shipped then the unused Resin and Filter and LD Hardware sheets shall remain blank.

Step 1: The user shall ensure that the conditions are met for each content type to be shipped in the RT-100 by checking the boxes of the respective loading table sheet (on top of each loading table sheet). If a mixture of resins/filters and activated hardware will be shipped, then the conditions of both loading sheets must be checked. Note that the conditions of the applicable contents must be checked before filling out the loading table and if any of the conditions is not satisfied, then the contents may not be shipped in the RT-100. One of the conditions is to verify that the bulk density of the resin or filter media does not exceed 1 g/cm3 at any point of the content and/or the density of the activated hardware is within the low-density hardware (2.0 g/cm3 < 7.5 g/cm3) or high-density hardware (7.5 g/cm3 9.0 g/cm3) range.

Note: Unchecked boxes does not prevent the user from using the loading table. The user is responsible for ensuring that the conditions are met and the boxes are checked.

Step 2: The total waste inventory volume and mass of each content type are entered in cubic feet and kilograms into the first two rows of each loading table sheet (in the top two yellow cells). Note that the mass cell will turn red if the mass input exceeds the maximum allowable mass for each content type. The maximum payload is limited to 6804 kg including contents, secondary containers, and shoring. The maximum activated hardware content mass is limited to 5,896 kg.

The waste volume is limited to 130 ft3 if the simplified loading curve is used for resins/filters or mixture of resins/filters and activated hardware shipments.

Step 3: The radionuclides in the waste stream are entered in the yellow column under Actual Content Nuclide. The user can type in the nuclide symbol (not case-sensitive) followed by the mass number without a space, dash, or hyphen (e.g., co60, ag100m, or CS137) or select the radionuclides from the drop-down menu in the cells. There are 200 empty cells on each loading table sheet, unused cells must be left blank. Note that the Maximum Allowable Activity Concentration (mCi/kg) will populate automatically in the second column once a radionuclide is entered into the table. The Maximum Allowable Activity Concentration (mCi/kg) limits of the 8 primary isotopes (ag110m, cs137, co58, co60, cs134, fe59, mn54, zn65) are mass dependent (higher limits with reduced waste mass). If N/A appears in any of the columns for the isotope entered, the isotope is not applicable to the RT-100 Loading Table or it is incorrectly input into the spreadsheet.17 17 If N/A appears in containment column but the isotope passes the shielding evaluation, the isotope is not in Table A-1 of Appendix A of 10 CFR 71. Look up the radiation type of the isotope. Based on Table A-3 of Appendix A of Robatel Technologies, LLC Page 7-35

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Step 4: The Actual Content Activity Concentration (mCi/kg) are entered in the third column for each radionuclide. The activity concentration input for each isotope should be the maximum for any waste stream of each conent shipped in the RT-100 cask. The weight of water must be excluded when determining the Ci/gram of radioactive content limit for resins and filters. Once the waste mass, radionuclides, and total activity concentrations are entered into the table, the rest is automatically updated (columns 4 through 10). For mixed filters and high-density hardware shipment as an example, the user shall input the maximum activity concentration in mCi/kg of each isotope for the filter content on the Filters & Resins sheet, and the maximum activity concentration of each isotope for the hardware content on the HD Hardware sheet.

Step 5: If there are multiple content types in the same package, the loading tables in the individual sheets are filled out based on the quantities (i.e., volumes, masses, and activity concentrations of radionuclides) for each of the contents separately. For example, if a package contains spent filters and activated velocity limiters (i.e., activated stainless steel), the Filters & Resins and HD Hardware loading table sheets are filled out for the two contents separately. Unused sheets must be left blank.

Step 6: The Loading Summary sheet is then checked to ensure that the combined contents (or one content type) have gamma sum less than 100%, A2 less than 3000, heat load less than 200 Watts, and neutron activity concentration less than 3.5 mCi/kg. Check the cells in the Loading Summary sheet beside Passed Criteria. If cell beside passed criteria is green and states TRUE, inventory has passed that particular set of criteria. If cell beside criteria is red and states FALSE, inventory has failed that particular set of criteria. An inventory must pass all test criteria (shielding, containment, heat load, and neutron limit) in order to be shipped in the RT-100 cask.

The pass/fail (TRUE/FALSE) of each test criteria shall be obtained from the Loading Summary sheet.

Note: There is an evaluation summary at the end of each loading sheet (Filters & Resins, LD Hardware, and HD Hardware) that shows the status of the each content type relative to the four test criteria. Its optional for the user to check these summaries, the same information can be obtained from the master Loading Summary sheet that must be checked by the user for each shipment.

Step 7: Check the Loading Curve sheet for the hydrogen gas generation status (applicable when hydrogenous materials are included). This is the simplified loading curve which is limited to 130 ft3 waste volume and 10-day shipping time and should be suitable for most shipments. If the Waste Volume (ft3) vs. Decay Heat (Watts) falls within the green region (acceptable region), then the shipment passes the hydrogen gas generation criterion for a 10-day shipping time. If a package 10 CFR 71, enter 0.54 Ci for A2 of beta or gamma emitters and 0.0024 Ci for alpha emitters.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 fails the simplified model criterion or if the waste volume is greater than 130 ft3, the shipping time is expected to exceed 10 days, or if any of the conditions in Table 4.4.4-1 are not met then the detailed hydrogen gas analysis must be performed.

Step 8: Below are the steps for performing the detailed hydrogen gas analysis in H2Gen Analysis sheet if necessary (less common option). The steps below are listed in the order the entry cells appear in H2Gen Analysis sheet.

1. Enter the air temperature (F) and atmospheric pressure (in Hg) at the time of loading

2. Either select the secondary container from the secondary container drop-down menu or enter the volume (ft3) occupied by the secondary container only (this volume shall not include waste volume and can be calculated based on the container weight and material density). Refer to Table 4.4.3-2 for the list of most commonly used containers, these should be available from the drop-down menu.

3. Enter the shoring volume (ft3), note that the shoring can be calculated using the shoring weight and material density.

4. Enter the Polyethylene or Polypropylene filter volume (ft3). Note that Polyethylene or Polypropylene filters have higher G Values than resins, therefore, they require performing the detailed analysis.

5. Enter the ionic resin bulk volume (ft3) which includes the volume of solid resin beads and the volume of the void space between the beads.

6. Enter steel filter volume (ft3).

7. Enter the hardware volume (ft3), this is the combined volume of low-density and high-density hardware in the waste.

8. Either select gamma/alpha decay heat distribution from the drop-down menu or enter the specific G-values based on the waste characterization. Refer to Table 7.5.2-2 for the G-values of expected hydrogenous materials based on the gamma/alpha decay heat distribution.

9. Either enter the decay heat of cask contents in watts to calculate the maximum allowable shipping time in days or enter the shipment time in days to calculate the maximum allowable decay heat in watts (one input is required).

7.6.1.3 Turkey Point Source Term Example Evaluation (Resin and Filter)

The following is a sample from stored spent resin and used filters in High Integrity Containers (HICs) from Turkey Point Units 3 & 4 Low Level Waste Facility [Ref. 9]. Sample results were increased by three standard deviations in order to generate a conservative source term, within 99%

confidence interval. Based on Reference 9, mass of a particular Turkey Point HICs contents was on the order of 2060 kg. Sample results were entered into the RT-100 Loading Table and the results are shown in Table 7.6.1-5 and Table 7.6.1-6.

The spent resin and used filter inventory from Turkey Point generates a value of 3.42% in the %

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 of Maximum column, which is below 100%. The inventory satisfies the dose rate regulatory compliance. The total inventory A2 value is 8.37 A2, which is below the 3000 A2 limit. The inventory satisfies the containment limit. The total power output is 1.13 Watts, which is below the 200 Watts limit. The actual content activity concentration of neutron emitters is 0.00 milliCuries/kilogram, which is below the limit of 3.5 milliCuries/kilogram. Therefore, this packages inventory would be considered acceptable.

Note that the 160 ft3 volume is used in this example for illustration purposes only and may not be the actual volume of this example package.

Table 7.6.1-5 Turkey Point Loading Table Example 160.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 2060.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) h3 5.83E+11 1.45E-03 0.00 1100 2.72E-06 3.37E-08 1.01E-07 N c14 5.83E+11 1.25E-03 0.00 81 3.18E-05 2.93E-07 7.55E-07 N fe55 1.63E+09 8.40E+00 0.00 1100 1.57E-02 3.40E-08 5.88E-04 N ni63 5.83E+11 4.68E+01 0.00 810 1.19E-01 1.02E-07 9.80E-03 N sr89 5.83E+11 1.73E-03 0.00 16 2.23E-04 3.46E-06 1.23E-05 N sr90 3.64E+08 1.46E-02 0.00 8.1 3.72E-03 6.70E-06 2.02E-04 N tc99 5.83E+11 4.10E-03 0.00 24 3.52E-04 5.02E-07 4.24E-06 N pu241 3.67E+09 4.52E-03 0.00 1.6 5.82E-03 3.18E-08 2.96E-07 N i129 5.83E+11 4.40E-05 0.00 Unlimited 4.68E-07 4.24E-08 N mn54 1.65E+04 9.17E-01 0.01 27 7.00E-02 4.98E-06 9.41E-03 N co57 1.99E+04 2.23E-01 0.00 270 1.70E-03 8.52E-07 3.92E-04 N co58 6.96E+03 1.00E+00 0.01 27 7.64E-02 5.99E-06 1.23E-02 N co60 2.50E+02 7.28E+00 2.91 11 1.36E+00 1.54E-05 2.31E-01 N sb125 1.29E+04 5.19E-01 0.00 27 3.96E-02 3.16E-06 3.38E-03 N cs134 6.03E+03 2.69E+01 0.45 19 2.92E+00 1.02E-05 5.65E-01 N cs137 7.95E+04 2.84E+01 0.04 16 3.66E+00 5.04E-06 2.95E-01 N ce144 1.80E+05 1.88E-01 0.00 5.4 7.17E-02 7.99E-06 3.10E-03 N cm242 4.05E+08 2.84E-05 0.00 0.27 2.17E-04 3.65E-05 2.13E-06 Y 2.84E-05 pu238 1.62E+09 9.97E-05 0.00 0.027 7.61E-03 3.31E-05 6.81E-06 Y 9.97E-05 pu239 1.14E+08 5.47E-05 0.00 0.027 4.17E-03 3.11E-05 3.50E-06 Y 5.47E-05 am241 6.02E+07 8.45E-05 0.00 0.027 6.45E-03 3.34E-05 5.81E-06 Y 8.45E-05 cm243 3.16E+04 9.16E-05 0.00 0.027 6.99E-03 3.66E-05 6.91E-06 Y 9.16E-05 Table 7.6.1-6 Turkey Point Example Loading Summary Robatel Technologies, LLC Page 7-38

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6.1.4 St. Lucie Loading Table (Resin and Filter)

The following is source term information for a typical St. Lucie HIC source term [Ref. 10]. Two sets of sample results were increased by three standard deviations in order to generate a conservative source term, within 99% confidence interval. If multiple samples are taken from a shipment, the maximum activity concentration values for each isotope should be used in the RT-100 Loading Table. Based on Reference 10, mass of a particular St. Lucie HICs contents was on the order of 1950 kg. The maximum values for each isotope from the two samples were inputted into the RT-100 Loading Table and the results are shown in Table 7.6.1-7 and Table 7.6.1-8.

The spent resin and used filter inventory from St. Lucie generates a value of 9.78% in the % of Maximum column, which is below 100%. The inventory satisfies the dose rate regulatory compliance. The total inventory A2 value is 1.69 A2, which is below the 3000 A2 limit. The heat load is 0.23 Watts, which is below the 200 Watt limit. The actual content activity concentration of neutron emitters is 0.00 milliCuries/kilogram, which is below the limit of 3.5 milliCuries/kilogram.

Therefore, this packages inventory would be considered acceptable.

Note that the 160 ft3 volume is used in this example for illustration purposes only and may not be the actual volume of this example package.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-7 St. Lucie Loading Table Example 160.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 1950.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) h3 5.83E+11 3.90E-03 0.00 1100 6.91E-06 3.37E-08 2.57E-07 N c14 5.83E+11 1.55E-02 0.00 81 3.73E-04 2.93E-07 8.86E-06 N fe55 1.63E+09 3.65E+00 0.00 1100 6.47E-03 3.40E-08 2.42E-04 N ni63 5.83E+11 3.00E+01 0.00 810 7.22E-02 1.02E-07 5.94E-03 N sr89 5.83E+11 9.20E-03 0.00 16 1.12E-03 3.46E-06 6.20E-05 N sr90 3.64E+08 2.86E-03 0.00 8.1 6.89E-04 6.70E-06 3.74E-05 N tc99 5.83E+11 2.70E-03 0.00 24 2.19E-04 5.02E-07 2.64E-06 N pu241 3.67E+09 4.00E-03 0.00 1.6 4.88E-03 3.18E-08 2.48E-07 N i129 5.83E+11 2.45E-04 0.00 Unlimited 4.68E-07 2.23E-07 N cr51 1.98E+05 1.46E+00 0.00 810 3.51E-03 2.20E-07 6.27E-04 N mn54 1.65E+04 8.35E-01 0.01 27 6.03E-02 4.98E-06 8.11E-03 N co57 1.99E+04 3.21E-01 0.00 270 2.32E-03 8.52E-07 5.34E-04 N co58 6.96E+03 1.59E+00 0.02 27 1.14E-01 5.99E-06 1.85E-02 N fe59 7.11E+02 9.00E-02 0.01 24 7.31E-03 7.74E-06 1.36E-03 N co60 2.50E+02 3.92E+00 1.57 11 6.95E-01 1.54E-05 1.18E-01 N zn65 2.34E+03 1.28E-01 0.01 54 4.62E-03 3.50E-06 8.73E-04 N nb94 6.40E+02 4.20E-02 0.01 19 4.31E-03 1.02E-05 8.34E-04 N zr95 2.62E+03 1.21E-01 0.00 22 1.07E-02 5.04E-06 1.19E-03 N nb95 2.59E+03 2.15E-01 0.01 27 1.55E-02 4.80E-06 2.01E-03 N ru103 1.83E+04 1.37E-01 0.00 54 4.95E-03 3.33E-06 8.89E-04 N ru106 1.57E+03 2.90E-01 0.02 5.4 1.05E-01 9.65E-06 5.45E-03 N ag108m 2.12E+03 5.20E-02 0.00 19 5.34E-03 9.70E-06 9.83E-04 N ag110m 4.84E+02 1.30E-01 0.03 11 2.30E-02 1.67E-05 4.23E-03 N sb124 2.19E+01 3.60E-02 0.16 16 4.39E-03 1.33E-05 9.31E-04 N sb125 1.29E+04 3.32E-01 0.00 27 2.40E-02 3.16E-06 2.05E-03 N cs134 6.03E+03 4.20E-02 0.00 19 4.31E-03 1.02E-05 8.35E-04 N cs137 7.95E+04 1.72E-01 0.00 16 2.10E-02 5.04E-06 1.69E-03 N la140 1.63E+01 1.29E+00 7.93 11 2.29E-01 1.68E-05 4.22E-02 N ce141 4.14E+04 2.92E-01 0.00 16 3.56E-02 1.46E-06 8.33E-04 N ce144 1.80E+05 5.70E-01 0.00 5.4 2.06E-01 7.99E-06 8.88E-03 N am241 6.02E+07 4.10E-05 0.00 0.027 2.96E-03 3.34E-05 2.67E-06 Y 4.10E-05 cm242 4.05E+08 1.56E-05 0.00 0.27 1.13E-04 3.65E-05 1.11E-06 Y 1.56E-05 cm243 3.16E+04 2.44E-04 0.00 0.027 1.76E-02 3.66E-05 1.74E-05 Y 2.44E-04 pu238 1.62E+09 5.70E-05 0.00 0.027 4.12E-03 3.31E-05 3.68E-06 Y 5.70E-05 pu239 1.14E+08 5.90E-05 0.00 0.027 4.26E-03 3.11E-05 3.58E-06 Y 5.90E-05 Table 7.6.1-8 St. Lucie Example Loading Summary Robatel Technologies, LLC Page 7-40

(Non-Proprietary Version)

RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6.1.5 Mixed Shipment Example (Resin/filter and Activated Hardware)

This mixed resins/filters and high-density hardware waste example is for illustration purposes only, the masses, volumes, radionuclides, and activity concentrations were chosen arbitrary. Table 7.6.1-9 and Table 7.6.1-10 show the filled out Resins & Filter and HD Hardware sheets, respectively.

Table 7.6.1-9 Mixed Loading Example (resin/filter sheet)

Table 7.6.1-10 Mixed Loading Example (high-density hardware sheet)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-11 shows the Loading Summary sheet for this mixed shipment example. Notice the TRUE green cells next to the test criteria cells which indicate that the package meets all four acceptance criteria and is acceptable for shipment in the RT-100 Cask.

Table 7.6.1-11 Mixed Loading Example (loading summary sheet)

Figure 7.6.1-1 shows the loading curve for this mixed shipment example. The decay heat generated in this example is 9.69 watts which is within the acceptable region (green region). The detailed analysis (H2Gen Analysis) is not required since the package passed the simplified hydrogen gas criteria. In this example, hydrogen gas concentration will not reach the 5% by volume limit during the 10-day shipment timeframe.

Figure 7.6.1-1 Mixed Shipment Loading Curve (Simplified Model)

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.6.1.6 Additional Examples Additional examples have been generated to show the maximum Co-60 loading for the RT-100 mass restricted filter and resin shipments (arbitrary masses were chosen within the established mass bands). Table 7.6.1-12, Table 7.6.1-13, Table 7.6.1-14, and Table 7.6.1-15 show the maximum Co-60 inventory allowed due to shielding limits for each mass limit. The activity concentration is at 100% under the % of Maximum column.

Table 7.6.1-12 Maximum Co-60 Filter and Resin Loading (<500 lbs) 3 20.00 Filter & Resin Waste Volume (ft )

Actual Content Loading Evaluation 200.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) co60 5.18E+02 5.18E+02 100.00 11 9.43E+00 1.54E-05 1.60E+00 N Table 7.6.1-13 Maximum Co-60 Filter and Resin Loading (<1,000 lbs) 40.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 400.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) co60 3.62E+02 3.62E+02 100.00 11 1.32E+01 1.54E-05 2.24E+00 N Table 7.6.1-14 Maximum Co-60 Filter and Resin Loading (<1,500 lbs) 60.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 600.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) co60 3.12E+02 3.12E+02 100.00 11 1.70E+01 1.54E-05 2.89E+00 N Table 7.6.1-15 Maximum Co-60 Filter and Resin Loading (>1,500 lbs) 3 80.00 Filter & Resin Waste Volume (ft )

Actual Content Loading Evaluation 800.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) co60 2.50E+02 2.50E+02 100.00 11 1.82E+01 1.54E-05 3.08E+00 N Robatel Technologies, LLC Page 7-43

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-16 and Table 7.6.1-17 show the effect of inputting two radionuclides that individually would pass the compliance tests, but having both radionuclide quantities in the cask would generate a dose rate that would fail either NCT or HAC shielding conditions. This is illustrated by the red FALSE cell due to shielding sum is 100.73%, which is above the 100% limit. The inventory in Table 7.6.1-16 would not be acceptable in an RT-100 Transport Cask.

Table 7.6.1-16 Failed Loading Table Example 160.00 Filter & Resin Waste Volume (ft 3)

Actual Content Loading Evaluation 3000.00 Filter & Resin Waste Mass (kg)

Gamma Emitting Nuclides Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Actual Content Actual Content Actual Maximum Allowable Activity Activity Activity Content Concentration (milliCuries/kg) Concentration % of Activity(i) / Q Value Heat Load Neutron Concentration Nuclide (milliCuries/kg) Maximum A2 (Curie) A2(i) (Watts/mCi) (Watts) Emitter? (milliCuries/kg) co60 2.50E+02 2.33E+02 93.20 11 6.35E+01 1.54E-05 1.08E+01 N cs137 7.95E+04 6.00E+03 7.55 16 1.13E+03 5.04E-06 9.07E+01 N Table 7.6.1-17 Failed Loading Table Example (Loading Summary)

Waste Volume Waste Mass Content Loading Evaluation Summary Content Type (ft3) (kg) Shielding Evaluation Containment Evaluation Heat Load Evaluation Neutron Emitter Evaluation Filters & Resins 160 3000 100.75 1188.55 101.51 Actual Content 0.00 Gamma Sum Gamma Sum Heat Load Activity Low-density Hardware -- -- -- -- -- --

(%) (A2) (Watts) Concentration High-density Hardware -- -- -- -- -- (milliCuries/kg) --

Total 160.00 3000.0 Total 100.75 Total 1188.55 Total 101.51 Total 0.00 Limit (% of 95% of Limit (Neutron the NCT and HAC 100 Limit (A2) 3000 Limit (Watts) 200 3.50 Sum, mCi/kg) dose rate criteria)

Passed Passed Shielding Passed Heat Load Passed Neutron FALSE Containment TRUE TRUE TRUE Criteria Criteria Limit Criteria Criteria Similarly, the Co-60 limits for low-density and high-density activated hardware are a function of the waste mass. Table 7.6.1-18 shows mass restricted limits of the 8 explicitly analyzed isotopes for each content type.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-18 Radionuclide Activity Concentration Limits (8 Explicitly Analyzed)

Specific Activity Limit (Ci/g)

Radionuclide 500 lbs 1,000 lbs 1,500 lbs 15,000 lbs ag110m 1.004E-03 6.952E-04 5.861E-04 4.840E-04 cs137 1.840E-01 1.758E-01 1.679E-01 7.950E-02 co58 1.492E-02 1.027E-02 8.335E-03 6.960E-03 Filters & Resins co60 5.184E-04 3.624E-04 3.124E-04 2.500E-04 cs134 1.210E-02 8.574E-03 7.398E-03 6.030E-03 fe59 1.400E-03 9.939E-04 8.600E-04 7.110E-04 mn54 2.961E-02 2.159E-02 1.918E-02 1.650E-02 zn65 4.540E-03 3.202E-03 2.809E-03 2.340E-03 Specific Activity Limit (Ci/g)

Radionuclide 1,000 lbs 2,000 lbs 8,000 lbs 13,000 lbs ag110m 8.552E-04 6.236E-04 4.393E-04 4.223E-04 cs137 1.640E-01 1.577E-01 1.163E-01 1.096E-01 Low-density Hardware co58 1.295E-02 9.091E-03 6.358E-03 6.327E-03 (2.0<7.5 g/cm3) co60 4.463E-04 3.227E-04 2.305E-04 2.262E-04 cs134 1.035E-02 7.784E-03 5.158E-03 5.207E-03 fe59 1.229E-03 8.970E-04 6.447E-04 6.242E-04 mn54 2.807E-02 2.097E-02 1.545E-02 1.530E-02 zn65 4.023E-03 2.926E-03 2.115E-03 2.095E-03 Specific Activity Limit (Ci/g)

Radionuclide 1,000 lbs 2,000 lbs 8,000 lbs 13,000 lbs ag110m 1.472E-03 1.057E-03 5.715E-04 5.047E-04 cs137 1.697E-01 1.590E-01 1.202E-01 1.062E-01 High-density Hardware co58 2.166E-02 1.549E-02 8.565E-03 7.607E-03 (7.59.0 g/cm3) co60 7.739E-04 5.485E-04 3.108E-04 2.644E-04 cs134 1.818E-02 1.335E-02 7.122E-03 6.174E-03 fe59 2.167E-03 1.533E-03 8.715E-04 7.489E-04 mn54 4.056E-02 3.586E-02 2.018E-02 1.734E-02 zn65 7.116E-03 5.103E-03 2.785E-03 2.438E-03 Robatel Technologies, LLC Page 7-45

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Table 7.6.1-19 Radionuclide Activity Concentration Limits (Generic Energy Line Method)

Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition na24 3.38E-06 NCT bi208 3.76E-06 NCT cs144 5.27E-06 NCT y88 1.02E-05 NCT la140 1.63E-05 NCT sb124 2.19E-05 NCT eu156 2.26E-05 NCT sc48 2.67E-05 NCT la138 3.87E-05 NCT tb156 4.10E-05 NCT ag106m 4.45E-05 NCT lu169 4.55E-05 NCT na22 5.38E-05 NCT sb120m 5.51E-05 NCT i124 5.54E-05 NCT br82 5.93E-05 NCT lu172 6.68E-05 NCT ta182 6.93E-05 NCT ca47 7.23E-05 NCT sc46 7.66E-05 NCT te131m 7.96E-05 NCT eu152 8.35E-05 NCT as72 8.49E-05 NCT tm172 9.32E-05 NCT eu154 9.69E-05 NCT cs136 1.09E-04 NCT pm148 1.10E-04 NCT ge69 1.25E-04 NCT tb160 1.56E-04 NCT sn125 1.82E-04 NCT rh102m 1.88E-04 NCT gd147 1.93E-04 NCT tc96 2.23E-04 NCT sr83 2.25E-04 NCT k40 2.42E-04 NCT co60 Refer to Table 7.6.1-18 NCT as76 2.84E-04 NCT Robatel Technologies, LLC Page 7-46

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition nb92 3.41E-04 NCT np238 3.49E-04 NCT am240 4.09E-04 NCT tb158 4.32E-04 NCT pm148m 4.79E-04 NCT ag110m Refer to Table 7.6.1-18 NCT ho166m 5.45E-04 NCT pa232 5.96E-04 NCT sb126 6.14E-04 NCT nb94 6.40E-04 NCT rb84 6.41E-04 NCT tm168 6.65E-04 NCT pr144 6.75E-04 NCT fe59 Refer to Table 7.6.1-18 NCT rh99 7.83E-04 NCT tm165 8.21E-04 NCT zr89 1.18E-03 NCT rh102 1.23E-03 NCT as71 1.37E-03 NCT ru106 1.57E-03 NCT tc95m 1.59E-03 NCT rb86 1.82E-03 NCT tc98 1.97E-03 NCT ag108m 2.12E-03 NCT ho166 2.17E-03 NCT ir194m 2.25E-03 NCT zn65 Refer to Table 7.6.1-18 NCT cs132 2.34E-03 NCT lu171 2.56E-03 NCT nb95 2.59E-03 NCT zr95 2.62E-03 NCT tb153 2.83E-03 NCT te121m 3.15E-03 NCT ag105 3.32E-03 NCT sb127 3.61E-03 NCT sb122 3.62E-03 NCT pm144 3.14E-03 NCT cd115m 4.17E-03 NCT kr79 4.27E-03 NCT Robatel Technologies, LLC Page 7-47

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition pm146 4.35E-03 NCT gd149 4.88E-03 NCT i126 5.10E-03 NCT tc99m 2.24E-02 HAC cs134 Refer to Table 7.6.1-18 NCT os185 6.57E-03 NCT cu64 6.61E-03 NCT pm143 6.71E-03 NCT co58 Refer to Table 7.6.1-18 NCT as74 7.27E-03 NCT ba131 7.28E-03 NCT br77 9.13E-03 NCT ce143 9.23E-03 NCT pm151 1.27E-02 NCT lu176 1.12E-02 HAC se75 1.14E-02 HAC y87 1.16E-02 HAC cm241 1.16E-02 HAC sr85 1.15E-02 HAC te121 1.15E-02 HAC rb83 1.23E-02 HAC rh101 1.30E-02 HAC sb125 1.29E-02 HAC hf181 1.36E-02 HAC mn54 Refer to Table 7.6.1-18 NCT y91 1.80E-02 NCT i131 1.71E-02 HAC xe127 1.73E-02 HAC ru103 1.83E-02 HAC yb169 1.97E-02 HAC co57 1.99E-02 HAC au198 2.01E-02 HAC ru97 2.01E-02 HAC ta183 2.04E-02 HAC np239 2.04E-02 HAC ba133 2.04E-02 HAC sn123 2.46E-02 NCT zr88 2.05E-02 HAC zn72 2.06E-02 HAC Robatel Technologies, LLC Page 7-48

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition te132 2.17E-02 HAC cf249 2.18E-02 HAC bi210m 2.20E-02 HAC hf182 2.22E-02 HAC rh101m 2.23E-02 HAC sn117m 2.25E-02 HAC hf175 2.30E-02 HAC u235 2.32E-02 HAC te123m 2.37E-02 HAC hg203 2.45E-02 HAC cm247 2.45E-02 HAC cf251 2.48E-02 HAC ce139 2.50E-02 HAC pu246 2.54E-02 HAC mo99 9.03E-03 NCT sc47 2.93E-02 HAC cs129 2.97E-02 HAC ag110 3.68E-02 NCT in113m 3.11E-02 HAC u237 3.13E-02 HAC cm243 3.16E-02 HAC pa233 3.22E-02 HAC pt191 3.21E-02 HAC tb155 3.46E-02 HAC cd115 3.40E-02 HAC ba140 3.43E-02 HAC er172 3.72E-02 HAC lu177m 3.73E-02 HAC cm245 3.76E-02 HAC np236 3.88E-02 HAC cu67 4.00E-02 HAC in114m 4.87E-02 NCT au199 4.09E-02 HAC ce141 4.14E-02 HAC ga67 4.31E-02 HAC tm167 4.47E-02 HAC ar37 4.49E-02 HAC th227 5.64E-02 HAC ra223 5.77E-02 HAC Robatel Technologies, LLC Page 7-49

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition pu237 6.27E-02 HAC nd147 6.37E-02 HAC sn126 6.59E-02 HAC cs137 Refer to Table 7.6.1-18 HAC sm153 6.61E-02 HAC os191 6.88E-02 HAC te129m 8.36E-02 NCT nb95m 7.52E-02 HAC rh105 8.17E-02 HAC eu155 9.11E-02 HAC re189 9.10E-02 HAC gd153 9.29E-02 HAC th229 9.61E-02 HAC os193 1.08E-01 HAC lu177 1.12E-01 HAC hf172 1.20E-01 HAC ba135m 1.28E-01 HAC yb175 1.69E-01 HAC ce144 1.80E-01 HAC eu149 1.91E-01 HAC be7 1.93E-01 HAC cr51 1.98E-01 HAC xe133m 1.99E-01 HAC pa231 2.09E-01 HAC re186 2.10E-01 HAC ir192 2.36E-01 HAC ag111 2.47E-01 HAC pm149 3.21E-01 NCT xe129m 4.35E-01 HAC ra224 4.99E-01 HAC ac225 5.06E-01 HAC kr81 5.11E-01 HAC ra226 5.69E-01 HAC np237 5.86E-01 HAC pt195m 6.57E-01 HAC as77 7.12E-01 HAC xe131m 1.02E+00 HAC sn113 1.09E+00 HAC th231 1.60E+00 HAC Robatel Technologies, LLC Page 7-50

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition dy166 1.70E+00 HAC kr85 2.62E+00 HAC ni59 3.30E+00 NCT am243 2.81E+00 HAC w188 3.13E+00 HAC nb91 3.47E+00 HAC th228 3.85E+00 HAC es254 4.48E+00 HAC tb161 4.57E+00 HAC u230 6.10E+00 HAC te125m 7.07E+00 HAC th234 7.08E+00 HAC am242m 1.20E+01 HAC rn222 1.50E+01 HAC w181 1.72E+01 HAC la137 1.73E+01 HAC pt193m 1.78E+01 HAC es253 1.90E+01 HAC u232 2.54E+01 HAC u233 2.67E+01 HAC xe133 2.72E+01 HAC th230 2.84E+01 HAC te127m 3.73E+01 HAC ac227 4.82E+01 HAC u234 5.50E+01 HAC am241 6.02E+01 HAC po210 7.99E+01 NCT pd103 6.91E+01 HAC th232 7.27E+01 HAC cd113m 8.69E+01 HAC u236 9.99E+01 HAC w185 1.07E+02 HAC v49 1.07E+02 HAC pu239 1.14E+02 HAC cf252 1.34E+02 HAC pu236 1.44E+02 HAC u238 1.86E+02 HAC pu240 2.60E+02 HAC sr90 3.64E+02 NCT Robatel Technologies, LLC Page 7-51

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition cl36 3.35E+02 HAC cm242 4.05E+02 HAC ca41 4.10E+02 HAC sm145 5.73E+02 HAC pu242 6.96E+02 HAC pm147 7.00E+02 HAC cm244 8.69E+02 NCT er169 1.53E+03 HAC pu238 1.62E+03 NCT fe55 1.63E+03 HAC pu241 3.67E+03 HAC bi210 1.96E+04 HAC bk249 5.28E+04 HAC pr143 2.15E+05 NCT tc97 5.81E+05 HAC ca45 5.83E+05 HAC ge71 5.83E+05 HAC nb93m 5.83E+05 HAC mo93 5.83E+05 HAC tc97m 5.83E+05 HAC cd109 5.83E+05 HAC sn119m 5.83E+05 HAC sn121m 5.83E+05 HAC te123 5.83E+05 HAC i125 5.83E+05 HAC i129 5.83E+05 HAC cs131 5.83E+05 HAC pm145 5.83E+05 HAC sm151 5.83E+05 HAC tb157 5.83E+05 HAC dy159 5.83E+05 HAC tm170 5.83E+05 HAC tm171 5.83E+05 HAC os194 5.83E+05 HAC pt193 5.83E+05 HAC tl204 5.83E+05 HAC pb205 5.83E+05 HAC pb210 5.83E+05 HAC ra225 5.83E+05 HAC Robatel Technologies, LLC Page 7-52

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Project RT-100 Transport Cask Task Maximum Curies Per Gram Limits Gamma Emitting Radionuclides Radionuclide Max. Ci/g Condition ra228 5.83E+05 HAC np235 5.83E+05 HAC cm246 5.83E+05 HAC cm248 5.83E+05 HAC cf250 5.83E+05 HAC se72 5.83E+05 HAC as73 5.83E+05 HAC te118 5.83E+05 HAC sb119 5.83E+05 HAC nd140 5.83E+05 HAC yb166 5.83E+05 HAC h3 5.83E+05 HAC ni63 5.83E+05 HAC sr89 5.83E+05 HAC tc99 5.83E+05 HAC sn113m 5.83E+05 HAC am242 5.83E+05 HAC c14 5.83E+05 HAC Robatel Technologies, LLC Page 7-53

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.7 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2021 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL, and the following specific Sections:

71.31(c) 71.45 71.47 71.47(b)(1) 71.89 71.87 71.47(b-d) 71.35(c) 71.43(g)

3. U.S. Department of Transportation, Hazard Communications for Class 7 (Radioactive)

Materials - Package and Vehicle Contamination Limits, 49 CFR 173.443.

4. U.S. Department of Transportation, Empty Class 7 (Radioactive) Materials Packaging, 49 CFR 173.428.

5. ANSI N14.5-2014, "American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

6. U.S. Nuclear Regulatory Commission, REGULATORY GUIDE 7.9 - Standard format and content of Part 71 applications for approval of packages for radioactive material, dated March 2005

7. U.S. Nuclear Regulatory Commission, Guide for Preparing Operating Procedures for Shipping Packages, NUREG/CR-4775, July 1988.

8. U.S. Nuclear Regulatory Commission, 10 CFR Part 20--STANDARDS FOR PROTECTION AGAINST RADIATION, and the following specific Sections:

20.1101(b) 20.1906 20.1906(e),

20.1906(a)

9. LLW-11-026, Turkey Point Unit 3 & 4 Low Level Waste Facility PTN Source Term Information, P. A. Miktus, September 7, 2011.

10. LLW-09-002, St. Lucie Unit 1 & 2 Low Level Waste Facility RFI Response, B.

Bedford, December 2, 2009.

11. [Withdrawn]

12. NUREG/CR-6673, "Hydrogen Generation in TRU Waste Transportation Packages,"

Anderson, B., Sheaffer, M., & Fischer, L., Lawrence Livermore National Laboratory, Livermore, CA, May 2000.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

8. ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Initially, RT employs an Assessment Test program to meet the requirements of 10 CFR Part 71

[Ref. 2], Subpart G. The RT-100 Package Maintenance Program ensures the RT-100 cask meets its Certificate of Compliance requirements throughout the package service life. Both the acceptance tests and maintenance programs are conducted in accordance with the RT Quality Assurance Program [Ref. 1]. Figure 8-1 shows flow of information for acceptance test and maintenance programs.

Figure 8-1 Information Flow for the Acceptance Tests and Maintenance Program Review General Structural Thermal Information Evaluation Evaluation

Codes and

Codes and

Temperatures Standards Standards

Pressures

Dimensions and

Bolt Torque

Heat Transfer Tolerances

Pressure Test Features

Materials

Structural Component Tests Containment Shielding Criticality Operating Evaluation Evaluation Evaluation Procedures

Fabrication

Shielding Tests

Not Applicable

Periodic Testing Verification

Shielding Material

Replacement Leakage Rate Specifications Component Testing

Periodic Verification Leakage Rate Acceptance Tests and Maintenance Program Acceptance Tests Maintenance Program

Visual and

Components and

Pressure and

Components and Measurement Materials Structural Materials

Weld Examination

Shielding

Leakage

Thermal

Pressure and

Thermal

Miscellaneous Structural

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1 Acceptance Tests Requirements for the acceptance testing include the following criteria:

o The SAR identifies codes, standards, and provisions of the quality assurance program used for the acceptance testing of the packaging in accordance with 10 CFR 71.31(c) and 71.37(b) [Ref. 2]. The SAR is prepared and the RT-100 is fabricated in accordance with the RT Quality Assurance Program [Ref. 1], approved by the USNRC on 21 March 2012.

o The fabrication of the RT-100 is verified to be in accordance with the approved design and in accordance with 10 CFR 71.85(c) [Ref. 2].

o The RT-100 is inspected for cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce its effectiveness; inspection is conducted in accordance with 10 CFR 71.85(a) [Ref. 2].

o Prior to lead pouring, measure the average clearance annular gap between the inner and outer shell of the cask body. Using a gauge, verify that the annular gap is above 86 mm at every point between the two shells. After lead pouring, the maximum annular gap between the lead and the inner and outer shell shall not exceed 0.1 cm.

o Since the RT-100 maximum normal operating pressure exceeds 35 kPa (5 psi) gauge, prior to first use the cask will be tested at an internal pressure at least 150% of the maximum normal operating pressure to verify its ability to maintain structural integrity at that pressure in accordance with 10 CFR 71.85(b) [Ref. 2].

o The RT-100 is conspicuously and durably marked with its model number, serial number, gross weight, and a package identification number assigned by the NRC in accordance with 10 CFR 71.85(c) [Ref. 2].

o Robatel Technologies, LLC performs all tests required by the NRC in accordance with 10 CFR 71.93(b) [Ref. 2].

o RT-100 is fabricated in accordance with drawings provided in Chapter 1, Appendix 1.4.

General Notes o A pre-shipment leak test is performed before each shipment of Type B waste, per ANSI N14.5-2014 [Ref. 6], as mentioned in Table 4.3-1.

o All disassembled parts will be reassembled in accordance with requirement stated in Chapter 7. In particular, this requirement applies to bolt/nut tightening, and torques applied.

o Cleanliness of sealing surfaces is of highest priority during package disassembly and assembly. This requirement particularly applies to O-rings and seal surfaces.

o O-rings are replaced within a 12 month period of use in accordance with Regulatory Guide 7.9 [Ref. 3] and NUREG-1609 [Ref. 4].

- The RT-100 has two (2) O-rings associated with each of the primary lid, secondary lid, and quick-disconnect valve.

- Replaced O-rings are leak tested in accordance with Section 8.1.4.2 or 8.1.4.3.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1.1 Visual Inspections and Measurements Throughout the fabrication process, confirmation by visual inspection and measurement are required to verify that the RT-100 packaging dimensionally conforms to the drawings RT100 PE 1001-1, Rev. H and RT100 PE 1001-2, Rev. H provided in Chapter1, Appendix 1.4. In addition, the packaging is to be visually inspected for any adverse conditions in materials or fabrication that would prevent the package from being assembled or operated in accordance with requirements outlined in Chapter 7, or tested in accordance with the requirements of Chapter 8. Visual and non-destructive examination shall be performed by ASNT or COFREND certified inspectors.

8.1.2 Weld Examinations Containment boundary welds are identified in drawing RT100 PRS 1011, Rev. E in Chapter 1, Appendix 1.4. The following welds on this drawing are classified as containment boundary welds:

S.1011.01, S.1011.02, and S.1011.03. These welds are required to be inspected and meet the acceptance requirements of ASME Code,Section III, Division I, Subsection ND, Article ND-5000

[Ref. 5].

The weld maps RT100 PRS 1011, Rev. E, RT100 PRS 1013, Rev. C, RT100 PRS 1031, Rev. D and RT100 PRS 1032, Rev. D listed in Chapter 1, Appendix 1.4, provide the examination criteria for each weld. Radiographic testing, dye penetrant testing, and/or visual testing is performed in accordance with applicable ASME standards. The Containment Boundary welds are also inspected by radiographic examination. Non-destructive examination shall be performed by ASNT or COFREND certified inspectors.

8.1.3 Structural and Pressure Tests A pressure test of the containment system is performed as required by 10 CFR 71.85 [Ref. 2]. As described in Chapter 3, Section 3.3.2.5, Maximum Normal Operating Pressure for the RT-100 cavity is 182.71 kPa. Per 10 CFR 71.85(b) [Ref. 2], the containment system shall be tested at an internal pressure at least 50% higher than the actual maximum normal operating pressure, or 274 kPa. However, for conservatism, the minimum test pressure is set to 300 kPa. The hydrostatic test pressure is held for a minimum of 10 minutes. Afterward, the primary lid and secondary lid closures are examined for leakage.

Except from temporary connections, leaks are remedied, and the test and inspection are repeated.

After depressurization and draining, the cask cavity and seal areas are visually inspected for cracks and deformation. Any cracks or deformation are remedied, and the test and inspection repeated.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1.4 Leakage Tests Section 8.1.4 describes the leakage tests to be performed on the RT-100 prior to its initial use.

Refer to Section 8.3 and Table 8.3.1-1 for a summary of the leak test types.

Testing performed on the cask body containment boundaries during fabrication:

8.1.4.1 Cask Containment Boundary 8.1.4.1.1 Cask Body Leak Testing - Prior to Lead Pouring 8.1.4.1.2 Primary Lid Assembly Including Secondary Lid and Cover Plate -

Prior to Final Assembly Verification testing performed on the cask after final assembly:

8.1.4.2 Primary and Secondary Lid Containment O-Rings Helium Leak Testing 8.1.4.3 Quick Disconnect Valve Helium Leak Testing 8.1.4.4 Quick Disconnect Valve Cover Plate Containment O-Rings Helium Leak Testing Note Regarding Test Personnel Qualifications Detailed procedures following the instructions below are to be approved by personnel certified in ASNT NDT or COFREND Level III leak testing. The use of COFREND certified personnel instead of ASNT certified personnel is accepted for leakage testing for the RT-100, based on the equivalence note 102885 EQN 001 Rev. C [Ref. 12].

Note Regarding Test Duration For each helium test, the duration must be calculated by test personnel. The Test Duration is a function of the System Response Time and the Helium Penetration Time.

The System Response Time is defined as the time from admitting helium to a test assembly with a known leak, until the measured leakage rate increases to 2x10-7 cm3/s above background. For the Primary Lid O-ring, which has the longest response time, the time has been experimentally determined to be less than 20 seconds.

The Helium Penetration Time is defined as the time from admitting helium to a test assembly, until the measured leakage rate of helium gas permeating through the seal under test increases to a rate of 2x10-7 cm3/s above background. For the Vent Port Cover Plate O-ring, which has the shortest penetration time, the time has been experimentally determined to be approximately 5 minutes.

The Test Duration should be such that:

2 x < < 80%

40 < < 240 In order to meet the above criterion, the Test Duration is specified as approximately 2 minutes.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1.4.1 Cask Containment Boundary 8.1.4.1.1 Cask Body Leak Testing - Prior to Lead Pouring Testing of the cask body containment boundary is performed prior to the lead shield pour to allow access to all containment welds and base material. This test is conducted using a helium leak detector in accordance with ANSI N14.5-2014 table A1 test A.5.3 [Ref. 6] to demonstrate compliance with the leaktight criteria. Figure 8.1.4-1 shows a general diagram of the test apparatus.

Calibration of the helium detector is performed using an appropriate leak standard, in accordance with Section 10 of ASTM E-499 [Ref. 7] or equivalent.

Figure 8.1.4-1 Cask Body Containment Boundary Test Apparatus o Test Personnel Qualifications Test personnel shall be ASNT NDT or COFREND Level II certified in leakage testing.

o Frequency Cask body containment boundaries are tested only once during fabrication.

o Components to be tested The body containment boundary includes the inner shell, the cask forged bottom, and the upper flange.

o Testing Procedure

1. Assemble the cask body with a substitute sealed plate used in place of the cask lid.

Note: The material in contact with the cask must be chemically compatible with the cask body material (stainless steel) and the test gas (helium).

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023

2. Place the entire vessel in a bag taped on the outer surface of the upper flange, as shown in Figure 8.1.4-1.

3. Create a vacuum in the cask cavity (0.01 atm abs or less).

4. Fill the bag with helium to a partial pressure of at least 25% of the total gas pressure.

5. Measure the helium flow signal detected in the interspace. The test duration will be approximately 2 minutes as described in Section 8.1.4.

o Acceptance Criteria Refer to Table 8.3.1-1 and Table 8.3.1-2 for test acceptance criteria.

o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

8.1.4.1.2 Primary Lid Assembly Including Secondary Lid and Cover Plate - Prior to Final Assembly Testing of the Primary Lid Assembly is performed prior to final assembly of the Cask. This test is conducted using a helium leak detector in accordance with ANSI N14.5-2014 table A1 test A.5.3

[Ref. 6] to demonstrate compliance with the leaktight criteria. Figure 8.1.4-2 shows a general diagram of the test apparatus. Calibration of the helium detector is performed using an appropriate leak standard, in accordance with Section 10 of ASTM E-499 [Ref. 7] or equivalent.

Figure 8.1.4-2 Primary Lid Assembly Containment Boundary Test Apparatus o Test Personnel Qualifications Test personnel shall be ASNT NDT or COFREND Level II certified in leakage testing.

o Frequency Cask body containment boundaries are tested only once during fabrication.

o Components to be tested The Primary Lid Assembly Containment Boundary includes the Primary Lid, Secondary Lid, and Quick-Disconnect Valve Cover Plate.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Testing Procedure

1. Remove the Quick-Disconnect Valve.

2. Assemble the Primary Lid per Chapter 7, Section 7.1.2.3, Secondary Lid Replacement, and 7.1.2.4, Quick-Disconnect Valve Cover Plate Replacement. The bolts must be torqued to the specifications listed in Table 7.4.5-1.

3. Attach to the Primary Lid Assembly a control tool as shown in Figure 8.1.4-2.

Note: The material in contact with the lid assembly must be chemically compatible with the primary lid assembly material (stainless steel) and the test gas (helium).

4. A sealing cover is arranged above the primary lid assembly as shown in Figure 8.1.4-2, and sealed just below the closure bolt flange.

Note: The sealing cover should not be placed over the opening leading to the Quick-Disconnect Valve Cover Plate.

5. Create a vacuum in the control volume (0.01 atm abs or less).

6. Fill the bag with helium to a partial pressure of at least 25% of the total gas pressure.

7. Measure the helium flow signal detected in the control volume. The test duration will be approximately 2 minutes as described in Section 8.1.4.

o Acceptance Criteria Refer to Table 8.3.1-1 and Table 8.3.1-2 for test acceptance criteria.

o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

8.1.4.2 Primary and Secondary Lid Containment O-Rings Helium Leak Testing Verification of the primary and secondary lid containment boundaries is performed prior to its initial use, periodically every 12 months, and after maintenance. This test is conducted using a helium leak detector in accordance with ANSI N14.5-2014 table A1 test A.5.3 [Ref. 6] to demonstrate compliance with the leaktight criteria. Calibration of the helium detector is performed using an appropriate leak standard, in accordance with Section 10 of ASTM E-499 [Ref. 7] or equivalent.

o Test Personnel Qualifications Test personnel shall be ASNT NDT or COFREND Level II certified in leakage testing.

o Frequency Maintenance leakage rate testing shall be performed prior to returning a package to service following maintenance, repair (such as a weld repair), or replacement of components of a containment boundary.

o Components to be tested The components tested are the inner O-rings in the primary lid or the secondary lid.

o Testing Procedure

1. Assemble the cask lids per Chapter 7, Section 7.1.2.2, Primary Lid Replacement, and Robatel Technologies, LLC Page 8-7

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 7.1.2.3, Secondary Lid Replacement. The bolts must be torqued to the specifications listed in Table 7.4.5-1.

2. Remove the quick disconnect valve cover plate, per Chapter 7, Section 7.1.1.3.

3. Remove the leak test port plug on either the primary or secondary lid, whichever containment boundary is to be tested. Attach the vacuum pump and the leak detection equipment to the port.

4. Pull a vacuum in the O-ring interspace (0.01 atm abs or less).

5. Fill the internal cavity with helium through the vent port (min helium partial pressure 25% of total pressure).

6. Measure the helium flow signal detected in the interspace. The test duration will be approximately 2 minutes as described in Section 8.1.4.

o Acceptance Criteria Refer to Table 8.3.1-1 and Table 8.3.1-2 for test acceptance criteria.

o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

8.1.4.3 Quick Disconnect Valve Helium Leak Testing As described in Section 4.1.2, the RT-100 does not rely on any valve or pressure relief device to meet the containment requirements. The Quick Disconnect valve is protected by the vent port cover plate, as shown in Figure 4.1.2-1. Therefore, it is not necessary for the Quick Disconnect Valve to meet the ANSI N14.5-2014 leaktight criteria. However, as an additional safety precaution, the Quick Disconnect valve is leak tested as described below.

Verification of the quick disconnect valve is performed prior to its initial use, periodically every 12 months, and after maintenance. This test is conducted using a helium leak detector in accordance with ANSI N14.5-2014 table A1 test A.5.3 [Ref. 6]. Calibration of the helium detector is performed using an appropriate leak standard, in accordance with Section 10 of ASTM E-499 [Ref. 7] or equivalent.

o Test Personnel Qualifications Test personnel shall be ASNT NDT or COFREND Level II certified in leakage testing.

o Frequency Maintenance leakage rate testing shall be performed prior to returning a package to service following maintenance, repair (such as a weld repair), or replacement of components of a containment boundary.

o Components to be tested The component tested is the quick-disconnect valve.

o Testing Procedure

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 in Table 7.4.5-2.

2. Install a bag on the vent port hole under the primary lid as shown in Figure 8.1.4-3.

Note: Alternately, the primary lid may be assembled to the cask body without the secondary lid, and the cavity filled with helium via the primary lid opening as shown in Figure 8.1.4-4.

3. Remove the quick disconnect valve cover plate, per Chapter 7, Section 7.1.1.3.

4. Install a vacuum clutch over the vent port. Pull a vacuum of 0.01 atm abs or less.

5. Fill the bag (or alternatively the containment vessel) with helium (min helium partial pressure 25% of total pressure).

6. Measure the helium flow signal detected in the interspace. The test duration will be approximately 2 minutes as described in Section 8.1.4.

o Acceptance Criteria Refer to Table 8.3.1-1 for test acceptance criteria.

o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

Figure 8.1.4-3 Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve Robatel Technologies, LLC Page 8-9

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Figure 8.1.4-4 Alternate Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve 8.1.4.4 Quick Disconnect Valve Cover Plate Containment O-Rings Helium Leak Testing Verification of the quick disconnect valve cover plate containment boundary is prior to its initial use, periodically every 12 months, and after maintenance. This test is conducted using a helium leak detector in accordance with ANSI N14.5-2014 table A1 test A.5.3 [Ref. 6] to demonstrate compliance with the leaktight criteria. Calibration of the helium detector is performed using an appropriate leak standard, in accordance with Section 10 of ASTM E-499 [Ref. 7] or equivalent.

o Test Personnel Qualifications Test personnel shall be ASNT NDT or COFREND Level II certified in leakage testing.

o Frequency Maintenance leakage rate testing shall be performed prior to returning a package to service following maintenance, repair (such as a weld repair), or replacement of components of a containment boundary.

o Components to be tested The component tested is the inner O-ring seal in the quick disconnect valve cover plate.

o Testing Procedure

1. Install a bag on the vent port hole under the primary lid as shown in Figure 8.1.4-5.

Note: Alternately, the primary lid may be assembled to the cask body without the secondary lid, and the cavity filled with helium via the primary lid opening as shown in Figure 8.1.4-6.

2. Remove the quick disconnect valve.

3. Assemble the cover plate per Chapter 7, Section 7.1.2.4 Quick Disconnect Valve Cover Plate Replacement. The bolts must be torqued to the specifications listed in Table 7.4.5-1.

4. Remove the quick disconnect valve cover plate leak test plug. Attach the vacuum pump and the leak detection equipment to the port.

5. Pull a vacuum in the O-ring interspace (0.01 atm abs or less).

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6. Fill the bag (or alternatively the containment vessel) with helium (min helium partial pressure 25% of total pressure).

7. Measure the helium flow signal detected in the interspace. The test duration will be approximately 2 minutes as described in Section 8.1.4.

o Acceptance Criteria Refer to Table 8.3.1-1 and Table 8.3.1-2 for test acceptance criteria.

o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

Figure 8.1.4-5 Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve Cover Plate Figure 8.1.4-6 Alternate Test Apparatus for Measuring the Helium Leak Rate through the Quick Disconnect Valve Cover Plate Robatel Technologies, LLC Page 8-11

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1.5 Component and Material Tests The RT-100 is fabricated using industry wide procurement practices for the materials. All materials undergo extensive testing to meet the ASME standards of the material specifications. These specifications (including ASME standards) are part of the procurement process to ensure component materials meet or exceed nuclear industry practices.

To confirm that acceptance criteria are met, materials used for the fabrication of the RT-100 are procured in accordance with RT Quality Assurance Program [Ref. 1] standards. Materials not meeting these standards are replaced. Any materials replacing the originally purchased component parts are subject to the same testing to ensure conformance with specifications.

Where possible, materials for the RT-100 are procured in accordance with ASME/ASTM standards.

In certain cases materials may be procured to other standards, such as DIN or ISO. For these materials, a Commercial Grade Dedication (CGD) Plan is prepared to ensure the material meets all specifications critical to safety; such CGD Plans may include independent analyses to confirm supplier material specifications. The CGD Plan is prepared in accordance with RT Quality Assurance Program [Ref. 1] requirements. The following sections give additional information regarding CGD Plan characteristics:

8.1.5.1 Foam 8.1.5.2 O-Ring 8.1.5.3 Ceramic Paper 8.1.5.4 Fusible Plugs 8.1.5.5 Carbon Steel and Alloy Steel Fasteners 8.1.5.6 Stainless Steel Fasteners 8.1.5.7 Thread Inserts 8.1.5.8 Quick Disconnect Valve 8.1.5.1 Foam The impact limiter foam is procured from General Plastics (GP), series FR3700. The density tolerance is specified as +/-10%. GP provides in its documentation the mechanical characteristics of the foam. As GP is a NQA-1 company, the critical characteristics of the foam are not retested.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b) 8.1.5.4 Fusible Plugs The fusible plugs used for the RT-100 are made of polyethylene. The critical characteristic of the fusible plugs is the melting temperature, which shall be less than 160°C. This temperature is based on Robatel Industries experience with other material plugs.

The melting temperature requirement is based on the ability of the plug to vent the pyrolysis gases produced during decomposition of the polyurethane impact limiter foam in the hypothetical fire accident. Polyurethanes do not break down below 475 K, or 202°C [Ref. 10]. Since the melting temperature of the fusible plug is less than the decomposition temperature of the impact limiter foam, the gases will be vented.

Another thermoplastic material can be used provided the fusible plugs have a melting temperature below 160°C.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Proprietary Information Content Withheld Under 10 CFR 2.390(b) 8.1.5.8 Quick Disconnect Valve The quick disconnect valve used in the RT-100 is a Staubli HCB 08.1152/IC/JE, made of stainless steel, and supplied with an EPDM O-ring. The critical characteristic of the valve is its leakage rate.

As previously described in Section 8.1.4.3, when assembled, the quick disconnect valve is leak tested.

8.1.6 Shielding Tests The RT-100 is designed to provide sufficient shielding to meet or exceed NRC and DOT requirements for a Type B (U)-96 package. Specifically, the RT-100 design includes gamma radiation shielding to meet 10 CFR Part 71 [Ref. 2] during both NCT and HAC.

Shielding integrity of the RT-100 is tested using nondestructive examination techniques (e.g.,

gamma scanning method) to ensure there are neither direct radiation streams nor voids greater than the acceptance criteria in the lead shield annulus. The acceptance criterion is defined as follows:

measurements indicate that no lead layer is less than the minimum specified on the drawing. Any results not meeting this requirement are remedied, and the test and inspection are repeated.

To ensure that the minimum lead thickness is poured, prior to lead pouring, a gauge of 86 mm is utilized to verify that at any point, the distance between the cask inner and outer shell is at least 86 mm. See Appendix 8.3, Section 8.3.1 for further information regarding the lead pouring process.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.1.7 Thermal Tests No thermal acceptance testing is required for the RT-100. Refer to the thermal evaluation of the RT-100 described in Chapter 3, Section 3.3, Thermal Evaluation under Normal Conditions of Transport and 3.4 Thermal Evaluation under HAC of the SAR.

8.1.8 Miscellaneous Tests No miscellaneous tests are to be performed on the RT-100 package.

8.2 Maintenance Program The RT-100 is subjected to routine inspection and periodic maintenance to ensure its compliance with this SAR and standards required by the NRC. In addition, requirements of the RT Quality Assurance Program [Ref. 1] are employed to direct required maintenance periods. Defective items are replaced or remedied, and tested as appropriate.

o The SAR identifies codes, standards, and provisions of the quality assurance program used for the maintenance of the RT-100 in accordance with 10 CFR 71.31(c) [Ref. 2], and 71.37(b) [Ref. 2]. The RT Quality Assurance Program [Ref. 1] addresses all criteria of 10 CFR 71 [Ref. 2].

o The RT-100 is maintained in an unimpaired physical condition other than superficial defects in accordance with 10 CFR 71.87(b) [Ref. 2]. Any major changes to the RT-100 result in an evaluation for repair/ replacement of damaged parts.

o The RT-100 is not designed to transport fissile material in accordance with 10 CFR 71.87(g)

[Ref. 2] and thus, the presence of any moderator or neutron absorber in a fissile material package is not applicable.

o RT shall perform any and all tests deemed appropriate by the NRC in accordance with 10 CFR 71.93(b) [Ref. 2]. NRC is permitted to perform tests on the RT-100 that it deems necessary.

o A maintenance program is part of the operational procedures to ensure that the package performs as intended throughout its service life.

Any RT-100 that does not comply with the specifications and verifications of the SAR is taken out of service until the corrective action(s) have been completed. All corrective actions are reported to RT, NRC, and approved RT-100 Users.

8.2.1 Structural and Pressure Tests No routine or periodic structural or periodic testing will be performed on the RT-100 transportation cask.

The RT-100 lifting fixture shall be tested annually in accordance with ANSI N14.6 [Ref.8]

requirements to verify continuing compliance.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Each containment weld is inspected from the inner surface using dye penetrant methods to detect cracks. Any results not meeting the ASME Code,Section III, Division I, Subsection ND, Article ND-5000 [Ref. 5] requirement are remedied, and the test and inspection are repeated.

8.2.2 Leakage Tests Section 8.2.2 describes leakage tests to be performed on the RT-100 during its use. This section is subdivided into testing performed after annual inspection or maintenance, and testing performed prior to each shipment. Refer to Appendix 8.3, Section 8.3.1 and Table 8.3-1 for a summary of the leak test types.

Note: Procedures described below are approved by ASNT NDT or COFREND Level III certified personnel in leakage testing.

8.2.2.1 Periodic and Maintenance Leak Test Leak testing of the RT-100 must be performed after completion of annual inspection and after maintenance or repair. These tests are identical to those performed on the RT-100 prior to its initial use. Refer to the following applicable subsections of Section 8.1.4 for details:

8.1.4.2 Primary and Secondary Lid Containment O-Rings Helium Leak Testing 8.1.4.3 Quick Disconnect Valve Helium Leak Testing 8.1.4.4 Quick Disconnect Valve Cover Plate Containment O-Rings Helium Leak Testing 8.2.2.2 Pre-Shipment Leak Test - Gas Pressure Rise Option A pre-shipment leakage test is required before each shipment of Type B material quantities to verify proper integrity of the containment system. The following test method is a gas-pressure rise approach in accordance with ANSI N14.5-2014 table A1 test A.5.2 [Ref. 6]. Test equipment shall be calibrated and traceable to an appropriate standard.

Note: As an alternative to the pressure rise method, the pre-shipment leak test can be performed following the pressure drop method described in Section 8.2.2.3.

o Test Personnel Qualifications Leakage rate testing shall be performed by personnel that are qualified and certified in accordance with the requirements of SNT-TC-1A-2006.

o Frequency Testing is performed prior to each shipment of Type B material.

o Components to be tested The vent port cover plate O-ring seals, and the primary or secondary lid O-ring seals, depending on which lid was removed prior to content loading. In case the contents are loaded through the primary lid, the secondary lid and vent port cover plate can be leak Robatel Technologies, LLC Page 8-22

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 tested before the contents are loaded, in accordance with ALARA principles.

In case the contents are loaded through the secondary lid, the primary lid and vent port cover plate can be leak tested before the contents are loaded, in accordance with ALARA principles.

Caution: Users of the RT-100 shall be aware that containment boundary components (detailed in Figure 4.1.2-1) could have been opened during a prior shipment of Type A contents, but a pre-shipment leakage rate test might not have been performed. The user must verify that an unopened lid has been previously leak tested in accordance with the Certificate of Compliance. If this verification cannot be made, the appropriate containment boundary seal must be leak tested.

o Testing Procedure

1. Assemble the cask lids per Chapter 6, Section 7.1.2.2, Primary Lid Replacement, 7.1.2.3, Secondary Lid Replacement, and Section 7.1.2.4, Quick-Disconnect Valve Cover Plate Replacement. The bolts must be torqued to the specifications listed in Table 7.4.5-1.

2. Remove the applicable leak test port plug.

3. Ensure that the O-ring on the test manifold is in good condition and lubricated. Connect the vacuum pump test assembly to the appropriate test port. The test assembly should consist of a vacuum pump isolated by a valve, with a gauge indicating the system pressure.

4. Accurately determine and record the control volume. The control volume includes the volume of the interspace between the O-rings as given in Table 8.2.2-1, plus the volume associated with the measuring instrumentation manifold.

Table 8.2.2-1 Volume of the Interspaces between the O-rings Interspace Location Volume [cm³] Volume [m³]

Primary Lid 70.0 0.000070 Secondary Lid 35.0 0.000035 Quick-Disconnect Valve Cover Plate 3.5 0.0000035

5. Determine and record the pressure gauge resolution, p.

6. Measure and record the base metal temperature of the cask lid, Tamb.

Note: Test should be carried out, where possible, in isothermal conditions. Small temperature variations can lead to large pressure variations.

7. Calculate the minimum required test duration, Hmin, following the method described in the Acceptance Criteria section below.

8. Create a vacuum in the interspace between the O-rings.

Note: Absolute pressure of 30 ~ 90 mbar is recommended. Lower absolute pressure may lead to outgassing of the interspace surfaces requiring longer pumping time to reach a sensitivity of 10-3 refcc/sec.

Note: Vacuum conditioning of the O-ring seals may be required.

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9. Isolate the pump. Physically disconnect the pump and/or turn the pump off.

10. Wait for the vacuum pressure to stabilize, with an absolute pressure of 30 ~ 90 mbar before starting the test.

11. Record the start time, t1, and start pressure, P1.

12. After the test duration, H, record the end time, t2, and end pressure, P2. Ensure that the test duration is greater than the minimum required test duration, Hmin.

13. Calculate the pressure change during the test, P. Ensure that the pressure change is less than or equal to the pressure gauge resolution, p.

Note: This test procedure confirms functionality of the containment seal and the control seal simultaneously. In the event of test failure, either O-ring may be responsible for the leakage.

14. Replace the applicable leak test port plug.

o Acceptance Criteria The preshipment leakage rate test need not be more sensitive than 1x10-3 ref-cm3/s [Ref. 6],

as shown in Table 8.3.1-1. This corresponds to a minimum sensitivity, Smin, under standard conditions of 1.01x10-4 Pa-m3/s. The test is carried out by the pressure rise method. Using formula B.14 given in Annex B of ANSI N14.5-2014, the test duration, H, must be greater than the minimum required test duration, Hmin:

> =

where: H = actual test duration [s]

Hmin = minimum required test duration [s]18 Smin = minimum required sensitivity [1.01x10-4 Pa-m3/s]

p = minimum measurable pressure [Pa] for the test, or gauge resolution VC = control volume [m3]

Tstd = standard temperature [298 K]

Tamb = base metal temperature of the cask lid [K] measured during the test Over the calculated test duration, there can be no measurable pressure rise. I.e., the minimum measurable pressure rise, P, must be less than or equal to the gauge resolution, p:

= 2 1 where: P1 = pressure [Pa] at the start of the test, t1 being the start time [s]

P2 = pressure [Pa] at the end of the test, t2 being the end time [s]

P = change in pressure [Pa] during the test p = minimum measurable pressure [Pa] for the test, or gauge resolution 18 Regardless of gauge resolution, the minimum required test duration shall be at least 10 seconds.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

Note: The pre-shipment leak test is not required before a shipment if the contents meet the definition for low specific activity materials or surface contaminated objects as stated in 10 CFR 71.4 [Ref.

2] and also, meet the exemption standard for low specific activity materials or surface contaminated objects as stated in 10 CFR 71.14(b)(3)(i) [Ref. 2].

8.2.2.3 Pre-Shipment Leak Test - Gas Pressure Drop Option A pre-shipment leakage test is required before each shipment of Type B material quantities to verify proper integrity of the containment system. The following test method is a gas-pressure drop approach in accordance with ANSI N14.5-2014 table A1 test A.5.1 [Ref. 6]. Test equipment shall be calibrated and traceable to an appropriate standard.

Note: As an alternative to the pressure drop option, the pre-shipment leak test can be performed following the pressure rise method described in Section 8.2.2.2.

o Test Personnel Qualifications Leakage rate testing shall be performed by personnel that are qualified and certified in accordance with the requirements of SNT-TC-1A-2006.

o Frequency Testing is performed prior to each shipment of Type B material.

o Components to be tested The vent port cover plate O-ring seals, and the primary or secondary lid O-ring seals, depending on which lid was removed prior to content loading. In case the contents are loaded through the primary lid, the secondary lid and vent port cover plate can be leak tested before the contents are loaded, in accordance with ALARA principles.

In case the contents are loaded through the secondary lid, the primary lid and vent port cover plate can be leak tested before the contents are loaded, in accordance with ALARA principles.

Caution: Users of the RT-100 shall be aware that containment boundary components (detailed in Figure 4.1.2-1) could have been opened during a prior shipment of Type A contents, but a pre-shipment leakage rate test might not have been performed. The user must verify that an unopened lid has been previously leak tested in accordance with the Certificate of Compliance. If this verification cannot be made, the appropriate containment boundary seal must be leak tested.

o Testing Procedure

1. Assemble the cask lids per Chapter 6, Section 7.1.2.2, Primary Lid Replacement, 7.1.2.3, Secondary Lid Replacement, and Section 7.1.2.4, Quick-Disconnect Valve Robatel Technologies, LLC Page 8-25

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 Cover Plate Replacement. The bolts must be torqued to the specifications listed in Table 7.4.5-1.

2. Remove the applicable leak test port plug.

3. Ensure that the O-ring on the test manifold is in good condition and lubricated. Connect the pump test assembly to the appropriate test port. The test assembly should consist of a pump isolated by a valve, with a gauge indicating the system pressure.

4. Accurately determine and record the control volume. The control volume includes the volume of the interspace between the O-rings as given in Table 8.2.2-1, plus the volume associated with the measuring instrumentation manifold.

5. Determine and record the pressure gauge resolution, p.

6. Measure and record the base metal temperature of the cask lid, Tamb.

Note: Tests should take place at isothermal conditions, if at all possible, as temperature changes lead to corresponding pressure changes.

7. The minimum required test duration, Hmin, shall be calculated, following the method described in the Acceptance Criteria section below.

8. Pressurize the cavity in the interspace between the O-rings.

Note: Minimum absolute pressure of 1.67 atm [24.5 psia] is recommended.

Another pressure differential may be used provided the cask user converts the reference leak rate for the new test conditions in accordance with NUREG/CR-6847 Section 2.2.6. Refer to Section 4.3.2 for details regarding the conversion of equivalent air leakage rates.

9. Isolate the pump. Physically disconnect the pump and/or turn the pump off.

Wait for the pressure to stabilize, with an absolute pressure of 1.67 ~ 1.75 atm

[24.5 ~ 25.7 psia] before starting the test.

Note: Another pressure differential may be used provided the cask user converts the reference leak rate for the new test conditions in accordance with NUREG/CR-6847 Section 2.2.6. Refer to Section 4.3.2 for details regarding the conversion of equivalent air leakage rates.

10. Record the start time, t1, and the start pressure, P1.

11. After the test duration, H, record the end time, t2, and end pressure, P2. Ensure that the test duration is greater than the minimum required test duration, Hmin.

12. Calculate the pressure change during the test, P. Ensure that the pressure change is less than or equal to the pressure gauge resolution, p.

Note: This test procedure confirms functionality of the containment seal and the control seal simultaneously. In the event of test failure, either O-ring may be responsible for the leakage.

13. Replace the applicable leak test port plug.

o Acceptance Criteria The preshipment leakage rate test need not be more sensitive than 1x10-3 ref-cm3/s [Ref. 6],

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Pa-m3/s, when the pressure in the test cavity is 1.67 atm and atmospheric pressure is 1 atm. The test is carried out by the pressure drop method. Using formula B.14 given in Annex B of ANSI N14.5-2014, the test duration, H, must be greater than the minimum required test duration, Hmin:

> =

where: H = actual test duration [s]

Hmin = minimum required test duration [s]19 Smin = minimum required sensitivity [1.01x10-4 Pa-m3/s]

p = minimum measurable pressure [Pa] for the test, or gauge resolution VC = control volume [m3]

Tstd = standard temperature [298 K]

Tamb = base metal temperature of the cask lid [K] measured during the test Over the calculated test duration, there can be no measurable pressure rise. I.e., the minimum measurable pressure rise, P, must be less than or equal to the gauge resolution, p:

= 1 2 where: P1 = pressure [Pa] at the start of the test, t1 being the start time [s]

P2 = pressure [Pa] at the end of the test, t2 being the end time [s]

P = change in pressure [Pa] during the test p = minimum measurable pressure [Pa] for the test, or gauge resolution o Actions to be taken if test fails Any condition which results in leakage in excess of the maximum allowable leak rate is corrected and re-tested.

Note: The pre-shipment leak test is not required before a shipment if the contents meet the definition for low specific activity materials or surface contaminated objects as stated in 10 CFR 71.4 [Ref. 2] and also, meet the exemption standard for low specific activity materials or surface contaminated objects as stated in 10 CFR 71.14(b)(3)(i) [Ref. 2].

8.2.3 Component and Material Tests Section 8.2.3 describes periodic tests and replacement schedules for components. This section is subdivided into routine component inspection and annual component inspection.

8.2.3.1 Routine Component Inspection Maintenance during normal use is performed to ensure that the RT-100 continues to meet design specifications and functions. Each time the RT-100 goes through the cycle of loading and unloading, the following components are visually inspected:

19 Regardless of gauge resolution, the minimum required test duration shall be at least 10 seconds.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 o Fasteners: Inspect threaded studs, bolts, nuts, washers, secure pins, and thread inserts.

Clean and lubricate; replace as necessary.

o Subcomponents: Inspect the condition of the primary lid, secondary lid, quick disconnect valve cover plate, upper impact limiter and lower impact limiter.

o Welds: Inspect the condition of cask attachment ring welds and cask lifting pocket welds.

o Seals: Inspect the RT-100 seals and check maintenance records to ensure the seals are within the 12 month replacement period. If replacement is necessary, contact RT and perform a leakage rate test after seal replacement.

o Labels: Inspect and record the readability of the RT-100 labeling. Repair if necessary.

8.2.3.2 Annual Component Inspection Inspections, tests and maintenance are performed every twelve (12) months of cask service as required in accordance with the SAR and NRC compliance requirements. The following steps are performed to ensure all components are in proper working order:

1. The exterior surfaces of the cask are visually inspected for damage and the results of the survey are documented. The major components and items to be inspected include the following items:

o Upper and Lower impact limiters o RT-100 body o Condition of the fusible plugs in the impact limiters o Condition and readability of RT-100 markings

2. Following procedures in Chapter 7, the RT-100 is disassembled into its components:

o Upper and Lower impact limiters o Leak test port plugs o Primary and Secondary lids o Cask body o Quick disconnect valve cover plate o Primary lid, secondary lid, and o Quick disconnect valve vent port cover plate O-rings

3. Cask visible exterior surface welds and interior cavity welds are visually inspected for defects.

4. The primary lid, secondary lid and quick-disconnect valve cover plate sealing surfaces are cleaned.

5. New inner and outer containment boundary O-rings are installed according to the recommendation of NUREG-1609 [Ref. 4].

6. The primary and secondary lid bolts shall be replaced after 500 cycles based on cask operator records. One cycle is defined as when the bolts are installed and fully torqued.

8.2.4 Thermal Tests No thermal tests are required for the RT-100. At least every four (4) years, testing is performed on the cask body lifting elements (lifting pockets). At the same time, examination of the inner shell visible welds parts on the cask body is performed in addition to the periodic maintenance every twelve (12) months.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.2.5 Miscellaneous Tests Threaded inserts may be used to repair threaded bolts holes. At a minimum, each repaired bolt hole will be tested for proper installation by assembling the joint components where the insert is used and ensuring the bolt can be tightened to the required torque. Refer to Tables 7.4.5-1 and 7.4.5-2 for applicable torque requirements.

If a threaded hole for lifting components is repaired, a load test shall be performed. The affected component must be able to withstand a load equal to 150% of the maximum service load. Each threaded insert shall be visually inspected after testing to ensure that there is no visible damage or deformation to the insert.

RT does not envision any other miscellaneous test being required of the RT-100.

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.3 Appendix 8.3.1 Summary of Leak Test Requirements Table 8.3.1-1 RT-100 Leakage Test Types ANSI Component(s) to be N14.5 Max. Leak Min.

Section Test Frequency Test Gas tested table A1 Rate Sensitivity test type Inner Shell Only once during 8.1.4.1.1 A.5.3 Helium Containment Boundary fabrication Table 8.3.1-2 Primary Lid Assembly Only once during 8.1.4.1.2 A.5.3 Helium Containment Boundary fabrication Table 8.3.1-2 Performed prior to Inner O-ring seals in the initial use, 8.1.4.2 primary lid or the A.5.3 periodically every Helium secondary lid 12 months, and Table 8.3.1-2 after maintenance Performed prior to initial use, 1x10-3 8.1.4.3 Quick-disconnect valve A.5.3 periodically every Helium No Leakage ref*cm3/sec 12 months, and after maintenance Performed prior to Inner O-ring seal in the initial use, 8.1.4.4 quick disconnect valve A.5.3 periodically every Helium cover plate 12 months, and Table 8.3.1-2 after maintenance Vent port cover plate Prior to each O-ring seals, and the N/A 1x10-3 8.2.2.2 A.5.2 shipment of No Leakage primary or secondary (vacuum) ref*cm3/sec Type B material lid O-ring seals Vent port cover plate Prior to each O-ring seals, and the 1x10-3 8.2.2.3 A.5.1 shipment of Air No Leakage primary or secondary ref*cm3/sec Type B material lid O-ring seals Table 8.3.1-2 Allowable Helium Leakage Rates Helium Partial Max. Leak Rate1 Min. Sensitivity Pressure 0.25 atm 2.672E-08 cm3/sec 1.336 E-08 cm3/sec 0.45 atm 5.137E-08 cm3/sec 2.569 E-08 cm3/sec 0.65 atm 8.185E-08 cm3/sec 4.093 E-08 cm3/sec 0.85 atm 1.263E-07 cm3/sec 0.632 E-07 cm3/sec 1.00 atm 1.897E-07 cm3/sec 0.949 E-07 cm3/sec 1

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RT-100 Safety Analysis Report, Rev. 10 Docket No. 71-9365 June 20, 2023 8.4 References

1. Robatel Technologies, LLC, Quality Assurance Program for Packaging and Transportation of Radioactive Material, 10 CFR 71 Subpart H, Rev. 6, Dated January 08, 2022 and NRC Approved on March 21, 2012

2. U.S. Nuclear Regulatory Commission, 10 CFR Part 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

3. U.S. Nuclear Regulatory Commission, REGULATORY GUIDE 7.9 - Standard format and content of Part 71 applications for approval of packages for radioactive material, dated March 2005

4. U.S. Nuclear Regulatory Commission, Standard Review Plan for Transportation Packages for Radioactive Material, NUREG-1609, March 31, 1999

5. ASME Boiler & Pressure Vessel Code 2007 Edition,Section III - Division 1 - Subsection ND, "Class 3 Components", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

6. ANSI N14.5-2014, "American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

7. ASTM E 499-1996, "Standard Test Methods for Leaks Using the Mass Spectrometer Leak Detector in the Detector Probe Mode," ASTM International, 100 Barr Harbor Drive, West Conshohocken, PA, www.astm.org.

8. ANSI N14.6-1978, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10000 pounds (4500 kg) or More for Nuclear Materials," American National Standards Institute, Inc., 11 West 42nd Street, New York, NY, www.ansi.org.

9. ASME Boiler & Pressure Vessel Code 2007 Edition,Section III - Division 1 - Subsection NF, "Supports", The American Society of Mechanical Engineers, Three Park Avenue, New York, NY, www.asme.org.

10. SFPE Handbook of Fire Protection Engineering, "Thermal Decomposition of Polymers,"

C.L. Hirschler, M. Marvelo, Chapter 7 of 3rd Edition, NFPA, 1 Batterymarch Park, Quincy, MA, 2001, www.nfpa.org.

11. DOT/FAA/AR-MMPDS-01, Metallic Materials Propertied Development and Standardization (MMPDS), U.S. Department of Transportation - Federal Aviation Administration, Washington, DC, www.ntis.gov.

12. 102885 EQN 001, Rev C, Equivalence Table - ASNT / COFREND - Qualification and Certification of NDE Personnel (PROPRIETARY)

13. Recommended Practice No. SNT-TC-1A-2006: Personnel Qualification and Certification in Nondestructive Testing. Columbus, OH: The American Society for Nondestructive Testing, Inc.

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0.6 x 515 x 0.015 x 1000 = 4635 kN/m Yield Cask Allowable 0.6 x Scy x Tc x 1000 0.6 x 199 x 0.035 x 1000 cya =

32 0.52 3650 9.81 1 kN Fb

0.58 kN 0.58 3.0 1 MPa sv

18 All other quantities are defined in Section 2.6.5.1 0.52 857 9.81 1 kN Fb

0.24 kN 0.24 3.0 1 MPa sv

960 mm RT100 PE 1001-1, Rev. H Appendix 1.4, the primary lid weighs 3648 kg and the secondary lid weighs 857 kg. Thus, (3648+857) w

1920 mm The remaining terms are as previously defined. Thus, sin(90) 9118.5 Ff

4,167.8 kN E l t 3l K1

1325.6 kN and sin(90) 1325.6 Fts

0.6 kN/m where, N b E b D4b Kb

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