ML20100F295

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Tn Americas LLC - Submittal of Renewal Application for the TN-68 Dry Storage Cask - CoC 1027
ML20100F295
Person / Time
Site: 07201027
Issue date: 04/09/2020
From: Narayanan P
Orano USA, TN Americas LLC
To:
Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML20100F292 List:
References
E-56131
Download: ML20100F295 (192)
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Text

April 9, 2020 E-56131 Columbia Office U. S. Nuclear Regulatory Commission 7135 Minstrel Way Columbia, MD 21045 Attn: Document Control Desk Tel: (410) 910-6900 One White Flint North

@Orano_USA 11555 Rockville Pike Rockville, MD 20852

Subject:

Renewal Application for the TN-68 Dry Storage Cask - CoC 1027

Reference:

U.S. Nuclear Regulatory Commission, Certificate of Compliance for Spent Fuel Storage Casks, Certificate No. 1027, Amendment 0, May 30, 2000, Docket No. 72-1027 Pursuant to 10 CFR 72.240, TN Americas LLC (TN) hereby submits the enclosed application for renewal of the spent fuel storage cask Certificate of Compliance (CoC) No. 1027 for the TN-68 Dry Storage Cask. The TN-68 Dry Storage Cask CoC 1027, Amendment 0 was approved by the NRC on May 30, 2000 for storage of spent nuclear fuel by general licensees. The current expiration date for CoC 1027 is May 28, 2020.

TN Americas LLC, as the certificate holder of CoC 1027, is applying for a renewal of CoC 1027 for an additional 40 years, in accordance with 10 CFR 72.240(a). This application for CoC 1027 renewal includes the safety analysis report (SAR) information required by 10 CFR 72.240(c), including design bases information, time-limited aging analyses, and a description of the aging management program to be implemented.

This submittal includes proprietary information which may not be used for any purpose other than to support NRC staff review of the application. In accordance with 10 CFR 2.390, I am providing an affidavit (Enclosure 1) specifically requesting that you withhold this proprietary information from public disclosure. Accordingly, a public version of the renewal application is provided as Enclosure 3.

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

Document Control Desk E-56131 April 9, 2020 Page 2 of 2 Should you have any questions regarding this submittal, please do not hesitate to contact Mr. Douglas Yates at 434-832-3101, or by email at Douglas.Yates@orano.group.

Prakash Narayanan Chief Technical Officer

Enclosures:

1. Affidavit Pursuant to 10 CFR 2.390
2. Certificate of Compliance Renewal Application for the TN-68 Dry Storage Cask, Certificate of Compliance No. 1027 (Docket No. 72-1027), Revision 0 (Proprietary Version)
3. Certificate of Compliance Renewal Application for the TN-68 Dry Storage Cask, Certificate of Compliance No. 1027 (Docket No. 72-1027), Revision 0 (Public Version) cc: Christian Jacobs, NRC DFM Doug Yates, TN Licensing

Enclosure l to E-56131 AFFIDAVIT PURSUANT TO 10 CFR 2.390 TN Americas LLC )

State of Mary land ) SS.

County of Howard )

I, Prakash Narayanan, depose and say that I am the Chief Technical Officer of TN Americas LLC, duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of IO CFR 2.390 of the Commission' s regulations for withholding this information.

The information for which proprietary treatment is sought is contained in Enclosure 2 and is listed below:

  • Enclosure 2, Portions of various chapters and appendices of the Certificate of Compliance No. 1027 Renewal Application, Revision 0 This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

I) The information sought to be withheld from public disclosure involves operating experience with spent fuel storage system aging and details of spent fuel storage system aging management studies, associated with the renewal application for the TN-68 Dry Storage Cask, which are owned and have been held in confidence by TN Americas LLC.

2) The information is of a type customarily held in confidence by TN Americas LLC, and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.
3) Public disclosure of the information is likely to cause substantial harm to the competitive position of TN Americas LLC, because the information consists of operating experience with the aging degradation of TN-68 loaded casks, and details of spent fuel storage system aging management studies and planned aging management strategies associated with the renewal application for the TN-68 Dry Storage Cask, the application of which provide a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their product's position or impair the position of TN Americas LLC's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

Fmiher the deponent sayeth not.

Prakash Narayanan

  • 't Chief Technical Officer, TN Americas LLC Su~ and s me this 24"' day of March,.., 2_0_2_0_._ _ __ _ _ __ _ _,

7

~ RONDA JONES Notary Public NOTARY PUBLIC MONTGOMERY COUNTY My Commission Expires )2___;-1.!!_I cl>;:93 MARYLAN D MY COMMISSION EXPIRES OCT. 16, 2023 Page I of I

Renewal Application for the TN-32 Dry Storage Cask, Certificate of Compliance No. 1021 (Docket No. 72-1021), Revision 0 (Proprietary Version)

Withheld Pursuant to 10 CFR 2.390

Enclosure 3 to E-56131 Certificate of Compliance Renewal Application for the TN-68 Dry Storage Cask, Certificate of Compliance No. 1027 (Docket No. 72-1027),

Revision 0 (Public Version)

PUBLIC Certificate of Compliance Renewal Application for the TN-68 Dry Storage Cask Certificate of Compliance No. 1027 (Docket No. 72-1027)

Prepared by:

TN Americas LLC Columbia, Maryland Revision 0 April 2020

CoC 1027 Renewal Application Revision 0 / April 2020 Table of Contents CHAPTER 1 GENERAL INFORMATION

1.1 INTRODUCTION

.............................................................................................. 1-1 1.2 TN-68 DRY STORAGE CASK SYSTEM DESCRIPTION .......................... 1-2 1.2.1 General System Description ................................................................. 1-2 1.2.2 Principal Components of the TN-68 Dry Storage Cask System ........... 1-2 1.2.2.1 TN-68 Cask ......................................................................... 1-2 1.2.2.2 Spent Fuel Assemblies ........................................................ 1-2 1.2.2.3 Concrete Storage Pad .......................................................... 1-3 1.2.2.4 Auxiliary Equipment ........................................................... 1-3 1.2.2.5 Miscellaneous Equipment ................................................... 1-3

1.3 BACKGROUND

................................................................................................ 1-4 1.4 APPLICATION FORMAT AND CONTENT ................................................ 1-5

1.5 REFERENCES

(CHAPTER 1, GENERAL INFORMATION) .................... 1-7 CHAPTER 2 SCOPING EVALUATION

2.1 INTRODUCTION

.............................................................................................. 2-1 2.2 SCOPING EVALUATION PROCESS AND METHODOLOGY ............................................................................................ 2-2 2.3 RESULTS OF SCOPING EVALUATION ..................................................... 2-4 2.3.1 TN-68 Cask ........................................................................................... 2-4 2.3.2 Spent Fuel Assemblies .......................................................................... 2-4 2.3.3 ISFSI Storage Pad ................................................................................. 2-4 2.3.4 Auxiliary Equipment ............................................................................. 2-4 2.3.5 Miscellaneous Equipment ..................................................................... 2-5 2.4 SUBCOMPONENTS NOT WITHIN SCOPE OF COC 1027 LICENSE RENEWAL ...................................................................................... 2-6 2.4.1 Overpressure System ............................................................................ 2-6 2.4.2 Drain Tube and Hansen Couplings ....................................................... 2-6 2.4.3 External Paint ........................................................................................ 2-6 2.4.4 Protective Cover Bolts, Seal, Washers, and Threaded Inserts .............. 2-6 2.4.5 Top Neutron Shield Bolts and Washers ................................................ 2-7 2.4.6 Lid Alignment Pin and Shear Key ........................................................ 2-7 Page i

CoC 1027 Renewal Application Revision 0 / April 2020 2.4.7 Security Wires and Seals ...................................................................... 2-7 2.4.8 Pressure Relief Valve ............................................................................ 2-7 2.4.9 Fuel Pellets ............................................................................................ 2-7

2.5 REFERENCES

................................................................................................... 2-8 CHAPTER 3 AGING MANAGEMENT REVIEW

3.1 INTRODUCTION

.............................................................................................. 3-1 3.2 AGING MANAGEMENT REVIEW METHODOLOGY ............................. 3-2 3.2.1 Identification of Materials and Environments ...................................... 3-2 3.2.2 Identification of Aging Mechanisms and Aging Effects ...................... 3-4 3.2.3 Determination of the Activities Required to Manage the Effects of Aging ................................................................................................ 3-4

3.3 DESCRIPTION

OF TN-68 DRY STORAGE CASK SYSTEM.................... 3-6 3.3.1 Description of TN-68 Cask Subcomponents ........................................ 3-6 3.3.2 Description of Storage Pad Subcomponents ......................................... 3-7 3.3.3 Description of Spent Fuel Assemblies Subcomponents ....................... 3-7 3.4 AGING MANAGEMENT REVIEW OF MATERIAL/ENVIRONMENT........................................................................ 3-9 3.4.1 Materials Evaluated .............................................................................. 3-9 3.4.2 Environments for the TN-68 Dry Storage Cask System SSCs ............. 3-9 3.4.3 Aging Effects Requiring Management ............................................... 3-10 3.4.3.1 Aging Mechanism of Steel Material ................................. 3-10 3.4.3.2 Aging Mechanism of Stainless Steel Material .................. 3-16 3.4.3.3 Aging Mechanisms of Aluminum Material ...................... 3-22 3.4.3.4 Aging Mechanism of Nickel Alloy Material .................... 3-27 3.4.3.5 Aging Mechanism of Polymer Material ........................... 3-29 3.4.3.6 Aging Mechanism of Borated Aluminum Material .......... 3-31 3.4.3.7 Aging Mechanism of Concrete Material........................... 3-35 3.4.3.8 Aging Mechanism of Spent Fuel Assembly Cladding...... 3-44 3.4.3.9 Aging Mechanism of Spent Fuel Assembly Hardware Materials ........................................................................... 3-49 3.4.3.10 Summary of Aging Mechanism of Materials.................... 3-52 3.5 AGING MANAGEMENT REVIEW FOR TN-68 CASK............................ 3-53 3.6 AGING MANAGEMENT REVIEW FOR STORAGE PAD ...................... 3-54 Page ii

CoC 1027 Renewal Application Revision 0 / April 2020 3.7 AGING MANAGEMENT REVIEW FOR SPENT FUEL ASSEMBLIES .................................................................................................. 3-55 3.8 OPERATING EXPERIENCE REVIEW RESULTS - AGING EFFECTS IDENTIFICATION ...................................................................... 3-56

3.9 REFERENCES

................................................................................................. 3-57 APPENDIX 3A TLAA IDENTIFICATION AND DISPOSITION 3A.1 Introduction ........................................................................................ 3A-1 3A.2 Methodology for Identification and Disposition of TLAAs .............. 3A-2 3A.2.1 Identification of TLAAs .................................................. 3A-2 3A.2.2 Disposition of Identified TLAAs ..................................... 3A-2 3A.3 Identified TLAAs ............................................................................... 3A-4 3A.4 Disposition of Identified TLAAs ....................................................... 3A-5 3A.4.1 End of Life Cavity Pressure ............................................. 3A-5 3A.5 Summary of TLAA Identification and Disposition ........................... 3A-6 3A.6 Summary of Updated End of Life Cavity Pressure Analysis ............ 3A-7 3A.7 References (Appendix A, TLAA Identification and Disposition) ........................................................................................ 3A-8 APPENDIX 3B SUPPLEMENTAL EVALUATIONS 3B.1 Introduction .........................................................................................3B-1 APPENDIX 3C OPERATING EXPERIENCE REVIEW 3C.1 Introduction .........................................................................................3C-1 3C.2 Operating Experience Review Approach............................................3C-2 3C.3 Internal and Industry Condition Reports.............................................3C-3 3C.3.1 Review of Internal Condition Reports ..............................3C-3 3C.3.2 Review of CoC Users Issue Reports .................................3C-3 3C.3.3 Review of Industry Condition Reports .............................3C-3 3C.4 Relevant International and Non-Nuclear Operating Experience ........3C-4 3C.5 Previous ISFSI Inspection Results ......................................................3C-5 3C.5.1 Prairie Island ISFSI Pre-Application Inspection...............3C-5 3C.5.2 North Anna ISFSI Pre-Application Inspection .................3C-6 3C.6 Licensee Event Reports.......................................................................3C-8 3C.7 Vendor-Issued Safety Bulletins ..........................................................3C-9 Page iii

CoC 1027 Renewal Application Revision 0 / April 2020 3C.8 NRC Generic Communications ........................................................3C-10 3C.8.1 NRC Information Notices ...............................................3C-10 3C.8.2 NRC Bulletins .................................................................3C-10 3C.9 Updated Consensus Codes, Standards, or Guides ............................3C-12 3C.9.1 Aging Management Reviews ..........................................3C-12 3C.9.2 Aging Management Programs ........................................3C-12 3C.10 Applicable Industry Initiatives ..........................................................3C-13 3C.10.1 High Burnup Fuel Demonstration Project ......................3C-13 3C.10.2 Chloride-Induced Stress Corrosion Cracking .................3C-13 3C.11 Conclusion ........................................................................................3C-14 3C.12 References .........................................................................................3C-15 CHAPTER 4 AGING MANAGEMENT PROGRAMS

4.1 INTRODUCTION

.............................................................................................. 4-1 4.2 AGING MANAGEMENT PROGRAM ELEMENTS ................................... 4-2 4.3 TN-68 AGING MANAGEMENT PROGRAM ............................................... 4-4 4.3.1 TN-68 AMP - Scope of Program ......................................................... 4-4 4.3.2 TN-68 AMP - Preventive Actions ........................................................ 4-4 4.3.3 TN-68 AMP - Parameters Monitored or Inspected .............................. 4-4 4.3.4 TN-68 AMP - Detection of Aging Effects ........................................... 4-5 4.3.4.1 TN-68 AMP - Interseal Pressure Monitoring ..................... 4-5 4.3.4.2 TN-68 AMP - Radiation Monitoring ................................. 4-5 4.3.4.3 TN-68 AMP - Visual Inspections....................................... 4-6 4.3.5 TN-68 AMP - Monitoring and Trending.............................................. 4-6 4.3.6 TN-68 AMP - Acceptance Criteria ...................................................... 4-7 4.3.6.1 TN-68 AMP - Interseal Pressure ........................................ 4-7 4.3.6.2 TN-68 AMP - Radiation Monitoring ................................. 4-7 4.3.6.3 TN-68 AMP - Visual Inspections....................................... 4-7 4.3.7 TN-68 AMP - Corrective Actions ........................................................ 4-8 4.3.8 TN-68 AMP - Confirmation Process ................................................... 4-8 4.3.9 TN-68 AMP - Administrative Controls ............................................... 4-8 4.3.10 TN-68 AMP - Operating Experience ................................................... 4-8 4.4 STORAGE PAD AGING MANAGEMENT PROGRAM ........................... 4-10 Page iv

CoC 1027 Renewal Application Revision 0 / April 2020 4.4.1 Storage Pad AMP - Scope of Program ............................................... 4-10 4.4.2 Storage Pad AMP - Preventive Actions ............................................. 4-10 4.4.3 Storage Pad AMP - Parameters Monitored or Inspected ................... 4-10 4.4.4 Storage Pad AMP - Detection of Aging Effects ................................ 4-11 4.4.5 Storage Pad AMP - Monitoring and Trending ................................... 4-11 4.4.6 Storage Pad AMP - Acceptance Criteria ............................................ 4-12 4.4.7 Storage Pad AMP - Corrective Actions ............................................. 4-13 4.4.8 Storage Pad AMP - Confirmation Process ......................................... 4-13 4.4.9 Storage Pad AMP - Administrative Controls ..................................... 4-13 4.4.10 Storage Pad AMP - Operating Experience ......................................... 4-13 4.5 HIGH BURNUP FUEL AGING MANAGEMENT PROGRAM ...................................................................................................... 4-15 4.5.1 HBU Fuel AMP - Scope of Program ................................................. 4-15 4.5.2 HBU Fuel AMP - Preventive Actions ................................................ 4-15 4.5.3 HBU Fuel AMP - Parameters Monitored or Inspected ...................... 4-16 4.5.4 HBU Fuel AMP - Detection of Aging Effects ................................... 4-16 4.5.5 HBU Fuel AMP - Monitoring and Trending ...................................... 4-16 4.5.6 HBU Fuel AMP - Acceptance Criteria .............................................. 4-17 4.5.7 HBU Fuel AMP - Corrective Actions ................................................ 4-17 4.5.8 HBU Fuel AMP - Confirmation Process............................................ 4-17 4.5.9 HBU Fuel AMP - Administrative Controls ....................................... 4-18 4.5.10 HBU Fuel AMP - Operating Experience ........................................... 4-18

4.6 REFERENCES

................................................................................................. 4-19 ATTACHMENT A CHANGES TO THE COC 1027 UPDATED FINAL SAFETY ANALYSIS REPORT A.1 INTRODUCTION............................................................................................. A-1 ATTACHMENT B CHANGES TO COC 1027AND TECHNICAL SPECIFICATIONS B.1 DETAILS OF PROPOSED COC CHANGES ................................................B-1 B.2 DETAILS OF PROPOSED TECHNICAL SPECIFICATIONS CHANGES ..........................................................................................................B-3 Page v

CoC 1027 Renewal Application Revision 0 / April 2020 Acronyms and Abbreviations ACI American Concrete Institute AMA Aging Management Activity AMID Aging Management INPO Database AMP Aging Management Program AMR Aging Management Review ANSI/ASCE American National Standards Institute / American Society of Civil Engineers ASME American Society of Mechanical Engineers ASR Alkali-Silica Reaction B&PV Boiler and Pressure Vessel BWR Boiling Water Reactor CAR Corrective Action Report CFR Code of Federal Regulations CISCC Chloride Induced Stress Corrosion Cracking CO Confinement Intended Function CoC Certificate of Compliance CR Sub-Criticality Control Intended Function DEF Delayed Ettringite Formation DHC Delayed Hydride Cracking DOE Department of Energy DSC Dry Shielded Canister DSS Dry Storage System EPRI Electric Power Research Institute GALL Generic Aging Lessons Learned HAZ Heat Affected Zone HBU High Burnup HDRP High Burnup Fuel Dry Storage Cask Research and Development Project INPO Institute of Nuclear Power Operations ISFSI Independent Spent Fuel Storage Installation ISG Interim Staff Guidance ITS Important-to-Safety LRA License Renewal Application MAPS Managing Aging Processes in Storage MIC Microbiologically Influenced Corrosion NDE Non-Destructive Examination NEI Nuclear Energy Institute NITS Not Important-to-Safety Page vi

CoC 1027 Renewal Application Revision 0 / April 2020 OE Operating Experience PCMI Pellet-to-Cladding Mechanical Interaction PEO Period of Extended Operation QA Quality Assurance RH Relative Humidity RT Retrievability Intended Function SAR Safety Analysis Report SCC Stress Corrosion Cracking SFA Spent Fuel Assembly SH Radiation Shielding Intended Function SR Structural Integrity Intended Function SS Stainless Steel SSCs Structures, Systems and Components TH Heat Removal Capability Intended Function TLAA Time-Limited Aging Analysis TLD Thermoluminescent Dosimeter TN TN Americas LLC TNUG TN Users Group TS Technical Specifications UFSAR Updated Final Safety Analysis Report Page vii

CoC 1027 Renewal Application Revision 0 / April 2020 CHAPTER 1 GENERAL INFORMATION CONTENTS 1.1 Introduction .................................................................................................................... 1-1 1.2 TN-68 Dry Storage Cask System Description ............................................................. 1-2 1.2.1 General System Description ................................................................................ 1-2 1.2.2 Principal Components of the TN-68 Dry Storage Cask System .................................................................................................................. 1-2 1.2.2.1 TN-68 Cask ........................................................................................... 1-2 1.2.2.2 Spent Fuel Assemblies .......................................................................... 1-2 1.2.2.3 Concrete Storage Pad ............................................................................ 1-3 1.2.2.4 Auxiliary Equipment ............................................................................. 1-3 1.2.2.5 Miscellaneous Equipment ..................................................................... 1-3 1.3 Background .................................................................................................................... 1-4 1.4 Application Format and Content ................................................................................. 1-5 1.5 References (Chapter 1, General Information) ............................................................ 1-7 LIST OF TABLES Table 1-1 Listing of CoC 1027 Amendments ...................................................................... 1-8 Page 1-i

CoC 1027 Renewal Application Revision 0 / April 2020 1.1 Introduction The TN-68 dry storage cask system, Certificate of Compliance (CoC) No. 1027, Revision 0 [1-3] was approved by the U.S. Nuclear Regulatory Commission (NRC) effective May 30, 2000, for storage of spent nuclear fuel by general licensees. The expiration date for CoC 1027 is May 28, 2020. As the certificate holder of CoC 1027, TN Americas LLC (TN)1 is applying for renewal of CoC 1027 for a term of 40 years, in accordance with the 10 CFR 72.240(a) [1-1].

This application for CoC 1027 renewal includes the safety analysis report (SAR) information required by 10 CFR 72.240(c). The SAR content of this application is based on the guidance provided in NUREG-1927 [1-2].

In accordance with NUREG-1927 [1-2], this renewal application is based on the continuation of the approved design basis throughout the period of extended operation.

The identification and management of potential aging degradation mechanisms for different material/environment combinations is based on the guidance of NUREG-2214 [1-7] in support of this renewal application.

1 TN Americas LLC, formerly AREVA TN, and Transnuclear, Inc.

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CoC 1027 Renewal Application Revision 0 / April 2020 1.2 TN-68 Dry Storage Cask System Description 1.2.1 General System Description The TN-68 dry storage cask system is a cask-based system for the dry storage of irradiated spent fuel assemblies (SFAs) and consists of a containment vessel, bolted lid, vent and drain port covers, a thick gamma shield, and radial neutron shielding.

Additional structures, systems, and components (SSCs) include a concrete storage pad, and other fuel transfer and auxiliary equipment used to support cask loading and transfer operations.

The following paragraphs provide an overview of the TN-68 dry storage cask system.

A more complete system description, including supporting design basis, is contained in the UFSAR, Revision 9 [1-6].

1.2.2 Principal Components of the TN-68 Dry Storage Cask System 1.2.2.1 TN-68 Cask The TN-68 containment vessel is comprised of an inner shell, which is a welded steel cylinder with an integrally-welded steel bottom; a welded flange forging (with stainless steel weld overlay); a bolted steel lid; and vent and drain port covers, cover seals and cover bolts. The inner shell is surrounded by a thick gamma shield. Radial neutron shielding around the gamma shield wall is provided by borated polyester resin encased in aluminum. A steel outer shell surrounds the radial neutron shielding.

Additional neutron shielding is provided by a top neutron shield, which consists of a disc of polypropylene encased in steel that is bolted to the cask lid. Trunnions are provided at the cask upper and lower ends to permit cask movement and transport. A protective weather cover fits over the top of the cask. The cover is sealed with an elastomer O-ring.

The internal basket assembly is made of stainless steel cells joined by a fusion welding process to stainless steel plates. Above and below the plates are slotted borated aluminum or boron carbide/aluminum metal matrix composite plates, which form an egg-crate structure. The basket is supported laterally by aluminum rails. The fuel basket structure is designed to hold 68 fuel assemblies.

1.2.2.2 Spent Fuel Assemblies The TN-68 dry storage cask system is designed to store boiling water reactor (BWR) spent fuel assembly types as authorized contents per the associated technical specifications (TS) and applicable amendments.

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CoC 1027 Renewal Application Revision 0 / April 2020 1.2.2.3 Concrete Storage Pad The TN-68 cask is installed on a load-bearing foundation, which consists of a reinforced concrete pad on a subgrade suitable to support the loads. There are no structural connections or means to transfer shear between the TN-68 cask and the concrete pad.

1.2.2.4 Auxiliary Equipment Auxiliary equipment used to facilitate cask loading, draining, drying, inerting and sealing operations include, but are not limited to, special lifting devices, vertical transporter, and vacuum drying/helium leak test equipment.

1.2.2.5 Miscellaneous Equipment Miscellaneous independent spent fuel storage installation (ISFSI) equipment (e.g.,

ISFSI security fences and gates, lighting, lightning protection, communications, and monitoring equipment) are not part of the CoC 1027 storage system.

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CoC 1027 Renewal Application Revision 0 / April 2020

1.3 Background

The TN-68 dry storage cask system, originally approved in May 2000, has evolved with time via approval of amendments to the original CoC 1027. Each of these amendments was designed to accommodate the evolving needs of the industry to store spent BWR fuel.

A listing of the currently approved CoC 1027 amendments is provided in Table 1-1.

This table provides: (a) the effective date of the amendment, and (b) a brief description of the scope of each amendment.

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CoC 1027 Renewal Application Revision 0 / April 2020 1.4 Application Format and Content This application includes SAR information required by 10 CFR 72.240 (c) [1-1]. The format and content of this SAR information is consistent with the guidance contained in NUREG-1927 [1-2].

Chapter 1, General Information: Chapter 1 provides (1) a general description of the TN-68 dry storage cask system, (2) a discussion of CoC 1027 amendments, and (3) information on the format and content of this application.

Chapter 2, Scoping Evaluation: Chapter 2 provides a description of the methodology used to identify the SSCs of the TN-68 dry storage cask system that are within the scope of the renewal. This methodology is based on the process described in NUREG-1927 [1-2]. Chapter 2 also provides a summary of the results of the scoping evaluation based on Revision 9 of the UFSAR [1-6].

Chapter 3, Aging Management Review: Chapter 3 provides the methodology used for the aging management review (AMR) of the TN-68 dry storage cask system, based on the guidance provided in NUREG-1927 [1-2]. The AMR documented in Chapter 3 identifies the materials and environment for those SSCs and associated subcomponents determined to be within the renewal scope in Chapter 2. This is accomplished by reviewing the drawings and the design basis included in the current UFSAR [1-6], CoC Amendment 1 [1-4], and associated TS [1-5]. Once the component material/environment combinations are determined, a review is performed to identify credible aging degradation mechanisms for the different material/environment combinations. NUREG-2214 [1-7] (based on technical literature, related research, industry information, and existing operating experience) is used as a guide for the review. After the credible aging mechanisms and effects are identified, it is determined whether the effects can be managed via a time-limited aging analysis (TLAA) or will require an aging management program (AMP).

Appendix 3A, Time-Limited Aging Analyses: Appendix 3A identifies the calculations or analyses used to demonstrate that in-scope SSCs will maintain their intended safety function throughout an explicitly stated period of operation, i.e., TLAAs. For those TLAAs that would not remain valid through the period of extended operation, Appendix 3A provides a summary of the revised or updated analysis.

Appendix 3B, Supplemental Evaluations: Appendix 3B provides a summary of supplemental evaluations and calculations performed to support the AMR and/or an element in an AMP.

Appendix 3C, Operating Experience Review: Appendix 3C provides a summary of the operating experience review performed to support the AMR and AMPs.

Chapter 4, Aging Management Programs: Chapter 4 provides the AMPs credited for managing each of the identified aging effects for the in-scope SSCs of the TN-68 dry storage cask system. The purpose of an AMP is to ensure that no aging effects result in a loss of intended safety function of the SSCs that are within the scope of renewal, for Page 1-5

CoC 1027 Renewal Application Revision 0 / April 2020 the term of the renewal. Each of the AMPs consists of the ten elements called for in NUREG-1927 [1-2].

Attachment A: This attachment provides the recommended changes to the UFSAR for CoC 1027 renewal.

Attachment B: This attachment provides the recommended changes to the CoC and TS for CoC 1027 renewal.

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CoC 1027 Renewal Application Revision 0 / April 2020 1.5 References (Chapter 1, General Information) 1-1 Title 10 Code of Federal Regulations Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

1-2 U.S. Nuclear Regulatory Commission, NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Revision 1, June 2016.

1-3 U.S. Nuclear Regulatory Commission, Certificate of Compliance for Spent Fuel Storage Casks, Certificate No. 1027, Revision 0, May 30, 2000, Docket No. 72-1027.

1-4 U.S. Nuclear Regulatory Commission, Certificate of Compliance for Spent Fuel Storage Casks, Certificate No. 1027, Amendment No. 1, Effective October 30, 2007, Docket No.

72-1027.

1-5 U.S. Nuclear Regulatory Commission, CoC 1027 Appendix A, TN-68 Generic Technical Specifications, Amendment 1, October 30, 2007, Docket No. 72-1027.

1-6 TN-68 Dry Storage Cask Updated Final Safety Analysis Report, Revision 9, May 2018.

1-7 U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, July 2019.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table 1-1 Listing of CoC 1027 Amendments Amendment Amendment Description No. Effective Date Initial approval to store spent fuel in the TN-68 dry 0 05/30/00 storage cask system.

Approved several changes, included the allowable fuel burnup, minimum cooling times, decay heat, and fuel 1 10/30/07 enrichment. The amendment also approved damaged fuel as authorized contents of the cask and to reduce the cask spacing on the storage pad.

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CoC 1027 Renewal Application Revision 0 / April 2020 CHAPTER 2 SCOPING EVALUATION CONTENTS 2.1 Introduction .................................................................................................................... 2-1 2.2 Scoping Evaluation Process and Methodology............................................................ 2-2 2.3 Results of Scoping Evaluation....................................................................................... 2-4 2.3.1 TN-68 Cask .......................................................................................................... 2-4 2.3.2 Spent Fuel Assemblies ......................................................................................... 2-4 2.3.3 ISFSI Storage Pad ................................................................................................ 2-4 2.3.4 Auxiliary Equipment ............................................................................................ 2-4 2.3.5 Miscellaneous Equipment .................................................................................... 2-5 2.4 Subcomponents Not Within Scope of CoC 1027 License Renewal............................ 2-6 2.4.1 Overpressure System ........................................................................................... 2-6 2.4.2 Drain Tube and Hansen Couplings ...................................................................... 2-6 2.4.3 External Paint ....................................................................................................... 2-6 2.4.4 Protective Cover Bolts, Seal, Washers, and Threaded Inserts ................................................................................................................... 2-6 2.4.5 Top Neutron Shield Bolts and Washers ............................................................... 2-7 2.4.6 Lid Alignment Pin and Shear Key ....................................................................... 2-7 2.4.7 Security Wires and Seals ..................................................................................... 2-7 2.4.8 Pressure Relief Valve ........................................................................................... 2-7 2.4.9 Fuel Pellets ........................................................................................................... 2-7 2.5 References ....................................................................................................................... 2-8 LIST OF TABLES Table 2-1 Scoping Evaluation of TN-68 Dry Storage Cask System SSCs .......................... 2-9 Table 2-2 Scoping Evaluation for TN-68 Cask .................................................................. 2-10 Table 2-3 Scoping Evaluation for Spent Fuel Assembly Subcomponents ......................... 2-13 Page 2-i

CoC 1027 Renewal Application Revision 0 / April 2020 2.1 Introduction Chapter 2 describes the evaluation process and methodology used to identify the structures, systems, and components (SSCs) of the TN-68 dry storage cask system that are within the scope of renewal.

In accordance with the guidance contained in NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel [2-2], the first step of the renewal process is the performance of a scoping evaluation. The objective of the scoping evaluation is to identify the SSCs of the TN-68 dry storage cask system that are within the scope of renewal.

A description of the scoping process and methodology is provided in Section 2.2. The results of the scoping evaluation are provided in Section 2.3. The bases for excluding certain subcomponents are provided in Section 2.4.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.2 Scoping Evaluation Process and Methodology The scoping evaluation of the Certificate of Compliance (CoC) No. 1027 storage system is performed based on the process described in NUREG-1927 [2-2]. SSCs (and associated subcomponents) are considered to be within the scope of the renewal if they satisfy either of the following criteria:

Criterion 1:

The SSC (and associated subcomponents) is classified as important-to-safety (ITS) as it is relied on to perform one of the following functions (Code of Federal Regulations (CFR) 10 CFR 72.3):

a) Maintain the conditions required by the regulations or the CoC to store spent fuel safely.

b) Prevent damage to the spent fuel during handling and storage.

c) Provide reasonable assurance that spent fuel can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public.

These SSCs ensure that ITS functions are met for (1) confinement, (2) radiation shielding, (3) sub-criticality control, (4) heat-removal capability, (5) structural integrity, and (6) retrievability.

Criterion 2:

The SSC (and associated subcomponents) is classified as not important-to-safety (NITS), but according to the design basis, its failure could prevent fulfillment of a function that is ITS.

The retrievability safety function is based on the ability to remove a cask loaded with spent fuel assemblies from its storage location; i.e., option C in Revision 2 of Interim Staff Guidance (ISG) 2 [2-7].

In accordance with NUREG-1927 [2-2], the renewal is based on the continuation of the approved design bases throughout the period of extended operation. Accordingly, the sources of information reviewed for this scoping evaluation that describe the approved design basis and the intended safety functions of the SSCs (and associated subcomponents) are the following:

a) TN-68 Updated Final Safety Analysis Report (UFSAR) [2-1].

b) CoC 1027 Certificate and Technical Specifications (TS) for each amendment

[2-3 and 2-5].

c) Safety Evaluation Reports (SERs) for each amendment [2-4 and 2-6].

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CoC 1027 Renewal Application Revision 0 / April 2020 These documents were reviewed to determine those SSCs (and associated subcomponents) with safety functions that meet either Scoping Criterion 1 or Criterion 2, as defined above. Based on this review, those subcomponents that perform or support any of the identified intended safety functions are determined to be in-scope and require an aging management review (AMR). Those subcomponents that do not perform or support a safety function are excluded from further evaluation in the AMR.

The scope of the CoC 1027 renewal encompasses the initial approved application (Amendment 0 [2-3]) and subsequently approved Amendment 1 [2-5].

Amendment 0 of CoC 1027 [2-3] approved the design of the TN-68 cask to store up to 68 intact, unconsolidated boiling water reactor (BWR) fuel assemblies. Section 1.2 of the associated U.S. Nuclear Regulatory Commission (NRC) Safety Evaluation Report

[2-4] states: The drawings for the TN-68 associated with structures, systems, and components (SSCs) important to safety are contained in Section 1.5 of the SAR. A specific list of these components is noted on the parts list shown on drawing 972-70-2.

The applicant provided sufficiently detailed drawings regarding dimension, materials, and specifications to allow for a thorough evaluation of the entire system. These drawings and the intended safety functions of the TN-68 casks were not affected by the subsequent amendment. Therefore, the drawings in Section 1.5 (along with the design information in Sections 2 through 14) of Revision 9 of the UFSAR [2-1]

represent the approved design bases for the TN-68. The drawings in Section 1.5 of the UFSAR were used as the source of the TN-68 subcomponents (along with their material specifications and safety classification) for the scoping evaluation.

Amendment 1 of CoC 1027 [2-5] approved several changes, which included the allowable fuel burnup, minimum cooling times, decay heat, and fuel enrichment. The amendment also approved damaged fuel as authorized contents of the cask and to reduce the cask spacing on the storage pad. Other than the additions of drawings to accommodate damaged fuel, these changes do not affect or change the subcomponents (including materials and safety classification) of the TN-68 casks.

In summary, the drawings and design information in UFSAR Revision 9 [2-1]

represent the approved design bases of the SSCs used for the scoping evaluation.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.3 Results of Scoping Evaluation The results of the scoping evaluation are summarized in Table 2-1. The following subsections provide a discussion of the scoping evaluation performed for each SSC.

2.3.1 TN-68 Cask The TN-68 cask includes, but is not limited to, the internal basket assembly, confinement vessel, gamma shielding surrounding the confinement vessel, radial neutron shielding, overpressure system, and protective cover. The TN-68 cask is classified as ITS. Therefore, it satisfies Criterion 1 and is in-scope. The scoping evaluations of the subcomponents of the TN-68 are summarized in Table 2-2, which also lists the UFSAR drawings used to identify the TN-68 subcomponents, their material of construction, and their safety classification.

2.3.2 Spent Fuel Assemblies The subcomponents of the spent fuel assemblies have intended safety functions required to maintain the conditions required by regulations to store the spent fuel safely. Therefore, the spent fuel assemblies satisfy Criterion 1 and are in-scope. The scoping evaluations of the subcomponents of the spent fuel assemblies are summarized in Table 2-3.

2.3.3 ISFSI Storage Pad The reinforced concrete storage pad is not classified as ITS in Section 2.3 of [2-1] and is, thus, considered a NITS structure. It is designed and constructed to plant-specific site conditions, and is not part of the CoC 1027 certification. However, a portion of the pad is included in the thermal models. Therefore, failure of the pad could affect the heat removal capability of the system, i.e., prevent fulfillment of an intended safety function. Therefore, the reinforced concrete storage pad meets Criterion 2 and is in-scope for renewal.

2.3.4 Auxiliary Equipment Auxiliary equipment used to facilitate cask loading, draining, drying, inerting and sealing operations include, but are not limited to, special lifting devices, vertical cask transporter, and vacuum drying/helium leak test equipment.

Part 1(b) of the CoC 1027 [2-3 and 2-5] states; The auxiliary equipment necessary for ISFSI operation is not included as part of the TN-68 cask system reviewed for a Certificate of Compliance under 10 CFR Part 72, Subpart L. Such equipment may include, but is not limited to, special lifting devices, transfer trailers or equipment, and vacuum drying/helium leak test equipment. Therefore, this equipment is not within the scope of CoC 1027 renewal.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.3.5 Miscellaneous Equipment Miscellaneous independent spent fuel storage installation (ISFSI) equipment (e.g.,

ISFSI security fences and gates, lighting, lightning protection, communications, and monitoring equipment) are not part of the CoC 1027 storage system approved in accordance with 10 CFR Part 72, Subpart L. They are not classified as ITS, nor would their failure prevent the fulfillment of a function that is ITS. Therefore, they are not within the scope of the CoC 1027 license renewal.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.4 Subcomponents Not Within Scope of CoC 1027 License Renewal This section provides a summary SSC subcomponents that were identified as not within the scope of renewal and the basis for exclusion.

2.4.1 Overpressure System Pressure monitoring equipment includes pressure switches or transducers and electrical cables. If the monitoring system was not to function, no safety functions of the cask would be impaired. There would be no leakage in or out of the cask. The overpressure system (including monitoring instrumentation) is designated as NITS (Table 2.3-1 of [2-1]) since the failure of the system will not result in a release of radioactive material. The monitoring system has not been designed to prevent failure during accident loadings. If an accident were to occur, measures would be taken to replace or repair the system soon after the accident. Therefore, the subcomponents of the overpressure system are not within the scope of CoC 1027 renewal.

2.4.2 Drain Tube and Hansen Couplings Drain tube (with all associated hardware including drain tube clamp, drain tube adapter, attachment screws, and O-ring seals) is for operational convenience only and does not perform any safety functions. The Hansen coupling on the vent and drain ports are for operational purposes only and do not perform any safety functions.

Failure of these components while in storage will not affect any safety functions.

Therefore, they are not within the scope of CoC 1027 renewal.

2.4.3 External Paint While a suitable primer and white topcoat paint is applied to the exterior of the cask to protect the cask from rusting, no credit for the coating is taken in the aging management review. Therefore, the external paint is not within the scope of CoC 1027 renewal.

2.4.4 Protective Cover Bolts, Seal, Washers, and Threaded Inserts The protective cover protects the overpressure tank, top neutron shield and lid from debris and wildlife nesting and allows rain water to drain more easily from the top of the cask. It has no structural integrity safety function. Operating experience has shown moisture can penetrate the protective cover. For the purpose of renewal, therefore, no credit is taken for the cover as a form of weather protection. The protective cover bolts, seal, flat washers, and threaded inserts are classified as NITS and since the protective cover is not being credited as a form of weather protection, failure of these subcomponents will not impact any intended safety functions. Therefore, the protective cover bolts, seal, flat washers, and threaded inserts are not within the scope of CoC 1027 renewal.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.4.5 Top Neutron Shield Bolts and Washers Per Section 2.3.1 of the UFSAR [2-1], the top neutron shield is used for supplemental shielding, but the accident condition dose limits are met without installing the top neutron shield. Thus, there is no structural integrity safety function for the top neutron shield, bolts or washers. Failure of the top neutron shield bolts or washers will not impact any intended safety functions. Therefore, the top neutron shield bolts and washers are not within the scope of CoC 1027 renewal.

2.4.6 Lid Alignment Pin and Shear Key Per Section 2.3.1 of the UFSAR [2-1], the lid alignment pin and shear key are used to ease operation and no structural credit is taken for these subcomponents. Failure of these NITS items will not affect any intended safety functions of the cask. Therefore, the lid alignment pin and shear key are not within the scope of CoC 1027 renewal.

2.4.7 Security Wires and Seals Per Section 2.3.1 of the UFSAR [2-1], the security wires and seals are used to provide evidence that the cask has not been tampered with. Therefore, failure of these NITS items will not impact any intended safety functions. Therefore, the security wires and seals are not within the scope of CoC 1027 renewal.

2.4.8 Pressure Relief Valve A pressure relief valve is provided in the outer shell to assure any pressure buildup due to heating of the resin and entrapped air is small. Per Table 3.3-6 of the UFSAR [2-1],

the primary function of the pressure relief valve is operations support. Failure of this NITS item will not impact any intended safety functions. Therefore, the pressure relief valve is not within the scope of CoC 1027 renewal.

2.4.9 Fuel Pellets The fuel pellets are stacked within the fuel cladding and do not perform a renewal intended function. The pellets may have sufficient cracking to have degraded into numerous pieces, rather than their original cylindrical shape. The pieces of pellets will be contained within the fuel cladding. It is the fuel cladding, not the fuel pellets that ensures the fuel remains in a subcritical and coolable geometry. Therefore, the fuel pellets are not within the scope of CoC 1027 renewal.

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CoC 1027 Renewal Application Revision 0 / April 2020 2.5 References 2-1 TN-68 Dry Storage Cask Updated Final Safety Analysis Report, Revision 9, May 2018.

2-2 U.S. Nuclear Regulatory Commission, NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Revision 1, June 2016.

2-3 U.S. Nuclear Regulatory Commission, Certificate of Compliance for Spent Fuel Storage Casks, Certificate No. 1027, Amendment 0, Effective Date May 30, 2000, Docket No.

72-1027, (ADAMS Accession Number ML003714545).

2-4 U.S. Nuclear Regulatory Commission, Safety Evaluation Report for Transnuclear, Inc.

TN-68 Dry Storage Cask System, Amendment 0, (ADAMS Accession Number ML003713477).

2-5 U.S. Nuclear Regulatory Commission, Certificate of Compliance for Spent Fuel Storage Casks, Certificate No. 1027, Amendment 1, Amendment Effective Date October 30, 2017, Docket No. 72-1027, (ADAMS Accession Number ML073050254).

2-6 U.S. Nuclear Regulatory Commission, Safety Evaluation Report for TN-68 Dry Storage Cask, CoC 1027 Amendment 1, Docket No. 72-1027, (ADAMS Accession Number ML073050286).

2-7 U.S. Nuclear Regulatory Commission, Division of Spent Fuel Management, Interim Staff Guidance 2, Fuel Retrievability in Spent Fuel Storage Applications, Revision 2, April 26, 2016.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table 2-1 Scoping Evaluation of TN-68 Dry Storage Cask System SSCs Intended Safety Function Criterion SSC Sub- Heat- In-Scope Radiation Structural Confinement Criticality Removal Retrievability No. 1 No. 2 Shielding Integrity Control Capability TN-68 Cask Yes Yes Yes Yes Yes Yes Yes N/A Yes Spent Fuel Assemblies Yes No Yes Yes Yes No Yes N/A Yes ISFSI Concrete Storage No No No Yes(1) No No No Yes Yes Pad Auxiliary Equipment No No No No No No No No No Miscellaneous No No No No No No No No No Equipment

1. Failure of the storage pad could prevent fulfillment of this intended safety function.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table 2-2 Scoping Evaluation for TN-68 Cask (3 Pages)

Intended Safety Function Criterion UFSAR Item Safety Sub- Heat-Rev Component Subcomponent Parts Radiation Structural Drawing No. No. Classification Confinement Criticality Removal Retrievability No. 1 No. 2 Shielding Integrity Control Capability 972-70-2 14 TN-68 1 Gamma Shield ITS No Yes No Yes Yes Yes Yes N/A Lid (including stainless steel 972-70-2 14 TN-68 2 ITS Yes Yes No Yes Yes No Yes N/A weld overlay at seal location) 972-70-2 14 TN-68 3 Inner Containment ITS Yes Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 4 Bottom ITS No Yes No Yes Yes Yes Yes N/A 972-70-2 14 TN-68 5 Bottom Containment ITS Yes Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 6 Upper Trunnion ITS No Yes No No Yes Yes Yes N/A 972-70-2 14 TN-68 7 Lower Trunnion ITS No Yes No No Yes Yes Yes N/A 972-70-2 14 TN-68 8 Shield Plate ITS No Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 9 Radial Neutron Shield ITS No Yes No Yes No No Yes N/A 972-70-2 14 TN-68 10 Outer Shell ITS No Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 11 Protective Cover NITS No Yes(2) No Yes(2) No No No Yes 972-70-2 14 TN-68 12 Top Neutron Shield NITS No Yes(2) No Yes(2) No No No Yes 972-70-2 14 TN-68 13 Radial N-Shield Box ITS No Yes No Yes No No Yes N/A 972-70-2 14 TN-68 14 Lid Bolt ITS Yes Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 15 Protective Cover Bolt NITS No No No No No No No No 972-70-2 14 TN-68 16 Lid Seal ITS Yes No No No No No Yes N/A 972-70-2 14 TN-68 17 Protective Cover Seal NITS No No No No No No No No (3) 972-70-2 14 TN-68 18 Overpressure Port Cover ITS Yes No No No No No Yes N/A Overpressure Port Cover 972-70-2 14 TN-68 19 ITS Yes(3) No No No No No Yes N/A Seal 972-70-2 14 TN-68 20 Top Neutron Shield Bolt NITS No No No No No No No No Pressure Monitoring 972-70-2 14 TN-68 21 NITS No No No No No No No No System 972-70-2 14 TN-68 22 Drain Port Cover ITS Yes Yes No Yes Yes No Yes N/A 972-70-2 14 TN-68 23 Vent Port Cover ITS Yes Yes No Yes Yes No Yes N/A Page 2-10

CoC 1027 Renewal Application Revision 0 / April 2020 Table 2-2 Scoping Evaluation for TN-68 Cask (3 Pages)

Intended Safety Function Criterion UFSAR Item Safety Sub- Heat-Rev Component Subcomponent Parts Radiation Structural Drawing No. No. Classification Confinement Criticality Removal Retrievability No. 1 No. 2 Shielding Integrity Control Capability Vent & Drain Port Cover 972-70-2 14 TN-68 24 ITS Yes No No No No No Yes N/A Seal Vent & Drain Port Cover 972-70-2 14 TN-68 25 ITS Yes No No No Yes No Yes N/A Bolts (SOC HD Cap)

Overpressure Port Cover 972-70-2 14 TN-68 26 ITS Yes(3) No No No Yes No Yes N/A Bolts (SOC HD Cap) 972-70-2 14 TN-68 27 Lid Alignment Pin NITS No No No No No No No No 972-70-2 14 TN-68 28 Basket Rail, Type 1 ITS No No Yes Yes Yes No Yes N/A 972-70-2 14 TN-68 29 Basket Rail, Type 2 ITS No No Yes Yes Yes No Yes N/A 972-70-2 14 TN-68 30 Basket Shim ITS No No Yes Yes Yes No Yes N/A 972-70-2 14 TN-68 31 Basket Rail Shim NITS No No No Yes(2) No No No Yes 972-70-2 14 TN-68 32 Fuel Compartment ITS No Yes Yes Yes Yes No Yes N/A 972-70-2 14 TN-68 33a(4) Poison Plate ITS No Yes Yes Yes No No Yes N/A 972-70-2 14 TN-68 33b(4) Aluminum Plate ITS No Yes Yes Yes No No Yes N/A 972-70-2 14 TN-68 34 Strucural Plates ITS No No Yes Yes Yes No Yes N/A Flange 972-70-2 14 TN-68 35 (including stainless steel ITS Yes Yes No Yes Yes No Yes N/A weld overlay) 972-70-2 14 TN-68 36 Shim ITS No Yes No No No No Yes N/A 972-70-2 14 TN-68 37 Trunnion Bolt ITS No No No No Yes Yes Yes N/A 972-70-2 14 TN-68 38 Deleted - - - - - - - - -

972-70-2 14 TN-68 39 Basket Hold Down ITS No Yes Yes No No No Yes N/A 972-70-2 14 TN-68 40 Shear Key NITS No No No No No No No No 972-70-2 14 TN-68 41 Pressure Relief Valve NITS No No No No No No No No 972-70-2 14 TN-68 42 Security Wire NITS No No No No No No No No 972-70-2 14 TN-68 43 Security Seal NITS No No No No No No No No 972-70-2 14 TN-68 44 Threaded Insert NITS No No No No No No No No 972-70-2 14 TN-68 45 Threaded Insert ITS Yes No No No Yes No Yes N/A 972-70-2 14 TN-68 46 Flat Washer NITS No No No No No No No No 972-70-2 14 TN-68 47 SOC Head Cap Screw ITS No Yes No No Yes No Yes N/A Page 2-11

CoC 1027 Renewal Application Revision 0 / April 2020 Table 2-2 Scoping Evaluation for TN-68 Cask (3 Pages)

Intended Safety Function Criterion UFSAR Item Safety Sub- Heat-Rev Component Subcomponent Parts Material Radiation Structural Drawing No. No. Classification Confinement Criticality Removal Retrievability No. 1 No. 2 Shielding Integrity Control Capability 972-70-2 14 TN-68 48 Shield Ring NITS No Yes(2) No No No No No Yes 972-70-2 14 TN-68 49 Flat Washer NITS No No No No No No No No N/A N/A TN-68 50 Not Used - - - - - - - - -

N/A N/A TN-68 51 Not Used - - - - - - - - -

Fuel Compartment 972-70-7 1 TN-68 52 ITS No No Yes No No No Yes No Extension 972-70-7 1 TN-68 53 End Cap, Top ITS No No Yes No No No Yes No 972-70-7 1 TN-68 54 End Cap Bottom ITS No No Yes No No No Yes No

1. The metallic seals consist of an inner spring, a lining, and a jacket. The spring is [ ] (a nickel alloy) or an equivalent material. The lining is stainless steel or nickel alloy. The jacket is made of aluminum.
2. Failure of this subcomponent could prevent fulfillment of this intended safety function.
3. The overpressure port cover, bolts, and seal are a barrier so that the pressure monitoring system is able to monitor the interseal pressure. Failure of the cover, bolts, or seal would not result in failure of the confinement boundary. However, they are classified as ITS and, thus, assigned a confinement function.
4. While Drawing 972-70-2 shows that Item #33 as Poison Plates, Note #7 on Drawing 972-70-5 says At the basket perimeter, Item 33 may be non-borated aluminum 1000 or 6000 series, as long as the pieces are retained in position by a positive means other than friction. To clearly identify that Item #33 could be two separate materials, Table 2-2 identified the poison plates as Item 33a and the optional aluminum plates as Item 33b.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table 2-3 Scoping Evaluation for Spent Fuel Assembly Subcomponents Intended Safety Function Criterion Material of Subcomponent Radiation Sub-Criticality Structural Heat-Removal Construction Confinement Retrievability No. 1 No. 2 Shielding Control Integrity Capability Fuel Pellets Uranium oxide No No No No No No No No Fuel Cladding and End Plugs Zircaloy Yes No Yes Yes Yes No Yes N/A Inconel Spacer Grid Assemblies No No Yes Yes No No Yes N/A Zircaloy Stainless Steel Upper End Fitting/Nozzle No No No Yes No No Yes N/A (and related subcomponents)

Inconel Lower End Fitting/Nozzle Stainless Steel (and related subcomponents) No No No Yes No No Yes N/A Inconel Water Channels Zircaloy No No Yes Yes No No Yes N/A Fuel Channel Zircaloy No No Yes No Yes No Yes N/A Page 2-13

CoC 1027 Renewal Application Revision 0 / April 2020 CHAPTER 3 AGING MANAGEMENT REVIEW CONTENTS 3.1 Introduction .................................................................................................................... 3-1 3.2 Aging Management Review Methodology ................................................................... 3-2 3.2.1 Identification of Materials and Environments ..................................................... 3-2 3.2.2 Identification of Aging Mechanisms and Aging Effects ..................................... 3-4 3.2.3 Determination of the Activities Required to Manage the Effects of Aging ................................................................................................... 3-4 3.3 Description of TN-68 Dry Storage Cask System ......................................................... 3-6 3.3.1 Description of TN-68 Cask Subcomponents ....................................................... 3-6 3.3.2 Description of Storage Pad Subcomponents ........................................................ 3-7 3.3.3 Description of Spent Fuel Assemblies Subcomponents ...................................... 3-7 3.4 Aging Management Review of Material/Environment .............................................. 3-9 3.4.1 Materials Evaluated ............................................................................................. 3-9 3.4.2 Environments for the TN-68 Dry Storage Cask System SSCs ..................................................................................................................... 3-9 3.4.3 Aging Effects Requiring Management .............................................................. 3-10 3.4.3.1 Aging Mechanism of Steel Material ................................................... 3-10 3.4.3.2 Aging Mechanism of Stainless Steel Material .................................... 3-16 3.4.3.3 Aging Mechanisms of Aluminum Material ........................................ 3-22 3.4.3.4 Aging Mechanism of Nickel Alloy Material ...................................... 3-27 3.4.3.5 Aging Mechanism of Polymer Material ............................................. 3-29 3.4.3.6 Aging Mechanism of Borated Aluminum Material ............................ 3-31 3.4.3.7 Aging Mechanism of Concrete Material............................................. 3-35 3.4.3.8 Aging Mechanism of Spent Fuel Assembly Cladding........................ 3-44 3.4.3.9 Aging Mechanism of Spent Fuel Assembly Hardware Materials ...... 3-49 3.4.3.10 Summary of Aging Mechanism of Materials...................................... 3-52 3.5 Aging Management Review for TN-68 Cask ............................................................. 3-53 3.6 Aging Management Review for Storage Pad............................................................. 3-54 3.7 Aging Management Review for Spent Fuel Assemblies ........................................... 3-55 3.8 Operating Experience Review Results - Aging Effects Identification .................... 3-56 Page 3-i

CoC 1027 Renewal Application Revision 0 / April 2020 3.9 References ..................................................................................................................... 3-57 Page 3-ii

CoC 1027 Renewal Application Revision 0 / April 2020 LIST OF TABLES Table 3-1 Approved Fuel Designs...................................................................................... 3-58 Table 3-2 Material Groupings ............................................................................................ 3-59 Table 3-3 Maximum TN-68 Cask Temperatures During Normal Storage ........................ 3-60 Table 3-4 Summary of Potential Aging Mechanisms ........................................................ 3-61 Table 3-5 Aging Management Review for TN-68 Cask .................................................... 3-65 Table 3-6 Aging Management Review Results Summary TN-68 Cask ............................ 3-69 Table 3-7 Aging Management Review for Storage pad ..................................................... 3-74 Table 3-8 Aging Management Review for Spent Fuel Assemblies ................................... 3-77 Page 3-iii

CoC 1027 Renewal Application Revision 0 / April 2020 3.1 Introduction This chapter describes the aging management review (AMR) of Certificate of Compliance (CoC) No. 1027 TN-68 dry storage cask system. The purpose of the AMR is to assess the need for aging management activities (AMA) for the structures, systems, and components (SSCs) determined to be within the scope of the CoC 1027 renewal. The AMR addresses aging mechanisms and effects that could adversely affect the ability of the SSCs to perform their intended safety functions during the period of extended operation (PEO).

Section 3.2 describes the AMR methodology, which follows the guidance and the processes of NUREG-1927 [3-1]. This section addresses each of the major steps of the AMR: Section 3.2.1 - Identification of Materials and Environments; Section 3.2.2 -

Identification of Aging Mechanisms and Aging Effects, and Section 3.2.3 -

Determination of the Activities Required to Manage the Effects of Aging.

Section 3.3 - Description of TN-68 Dry Storage Cask System, provides a brief description of the in-scope CoC 1027 SSCs.

Section 3.4 - Aging Management Review of Material/Environment, lists the materials of construction and the environments for the in-scope SSCs. The section also evaluates the potential aging degradation mechanisms/effects based on the material/environment combination. For each mechanism, a determination is made of whether the mechanism is considered credible in each environment that the material is exposed to during the PEO.

Sections 3.5, 3.6, and 3.7 provide the AMR results of the TN-68 cask, storage pad, and spent fuel assemblies (SFAs), respectively.

Section 3.8 provides a summary of the conclusions of the operating experience (OE) review contained in Appendix 3C.

Appendix 3A, Time-Limited Aging Analyses, identifies the calculations or analyses used to demonstrate that in-scope SSCs will maintain their intended safety function throughout an explicitly stated period of operation, i.e., time-limited aging analyses (TLAAs).

Appendix 3B, Supplemental Evaluations, provides a summary of supplemental evaluations and calculations performed to support the AMR and/or an element in an AMP.

Appendix 3C, Operating Experience Review, provides a summary of the operating experience review performed to support the AMR and aging management programs (AMPs).

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CoC 1027 Renewal Application Revision 0 / April 2020 3.2 Aging Management Review Methodology The AMR follows the methodology recommended in NUREG-1927 [3-1]. The AMR provides an assessment of the aging effects that could adversely affect the ability of the SSCs to perform their intended safety functions during the PEO.

The AMR process involves the following major steps:

  • Identification of materials and environments
  • Identification of aging effects and mechanisms requiring management
  • Determination of the activities required to manage the effects and mechanisms of aging; this involves the identification of TLAAs or AMPs for managing the effects of aging.

The scoping and screening evaluation in Chapter 2 identifies the in-scope SSCs for which potential aging effects must be identified and evaluated. For each SSC, the material of construction and the environment to which each SSC is exposed are determined. The component environments are determined based on the location of the component within the storage system. Once the component material/environment combinations are determined, a review of NUREG-2214 [3-2] is performed to identify credible aging degradation mechanisms for the different material/environment combinations. NUREG-2214 is based on technical literature, related research, industry information, and existing OE. After the credible aging mechanisms and effects are identified, it is determined whether the effects can be managed via a TLAA, or will require an AMP.

3.2.1 Identification of Materials and Environments The first step in the AMR process is to identify the materials of construction for each subcomponent of the in-scope SSCs and the environments to which those materials are exposed during normal storage conditions. The combinations of materials and environments are used to identify the potential aging effects that require management during the PEO.

Materials The TN-68 dry storage cask system SSCs and associated subcomponent materials of construction are summarized in:

CoC 1027 Renewal Application Revision 0 / April 2020 The materials of construction were identified through a review of the drawings provided in the Updated Final Safety Analysis Report (UFSAR) [3-3], along with other pertinent design information.

Environments The environments to which SSCs and associated subcomponents are exposed play a critical role in the determination of potential aging mechanisms and effects. A review of the information presented in Chapter 2 of the UFSAR [3-3] was performed to assess the environmental conditions to which the SSCs are normally exposed. The configuration of a TN-68 dry storage cask system at an independent spent fuel storage installation (ISFSI) consisting of an array of individual TN-68 casks designed to provide an effective means of protection against extreme seasonal weather conditions as described in Reference [3-3].

The TN-68 dry storage cask system has been designed and qualified for a wide range of environmental conditions. The cask temperature response to changes in ambient conditions will be relatively slow because the cask thermal inertia is large. A daily average minimum temperature of -20 °F and a daily average maximum temperature of 100 °F were used in the thermal evaluation of the cask.

The environments to which the TN-68 dry storage cask system is exposed are affected by the characteristics of the ISFSI site environment, as well as by the component location within the storage system. Six basic environments that apply for the TN-68 dry storage cask system SSCs and subcomponents:

  • Air-outdoor - In this environment, components are directly exposed to weather, including precipitation and wind. During storage, the exterior surfaces of the cask are exposed to all outdoor weather conditions, including insolation, wind, rain, snow, and site-specific ambient air conditions, including moist, possibly salt-laden atmospheric air, ambient temperatures, and humidity. Operating experience has shown moisture can penetrate the protective cover. Therefore, for the purpose of CoC renewal, no credit is taken for the protective cover as a form of weather protection.
  • Embedded-in-concrete - In this environment, one or more surfaces of a component are in contact with concrete, e.g., rebar. This may prevent ingress of water and contaminants to the embedded surface, depending on the permeability of the embedding environment.
  • Embedded-in-metal - In this environment, one or more surfaces of a component are in contact with another component or material. This may prevent ingress of water and contaminants to the embedded surface, depending on the permeability of the embedding environment. However, for the purposes of this AMR, the materials in this environment are treated as though they are exposed to the surrounding environment. For example, the surfaces of the poison plates are in contact with the fuel compartment and structural plates, but are treated as though they are exposed to the helium environment.

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CoC 1027 Renewal Application Revision 0 / April 2020

  • Fully encased - In this environment the component is fully enclosed inside another component, or the surface between two components is sealed, or fully lined by another material (e.g., steel), which prevents ingress of water and contaminants. An example is the radial neutron shield encased between the gamma shield and the outer shell.
  • Helium - In this environment, the component surfaces are exposed to the helium fill gas inside the cask and trace quantities of other gases, such as nitrogen, oxygen, argon, and fission product gases. This environment applies to the fuel pin, cladding, and other internal components inside the cask.
  • Groundwater/Soil - In this environment, the component surface is exposed to a soil environment, i.e., below grade. Groundwater is subsurface water found in wells, tunnels, or drainage galleries, or water that flows naturally to the earths surface via seeps or springs. Soil is a mixture of organic and inorganic materials produced by the weathering of rock and clay minerals or the decomposition of vegetation. Below-grade concrete structures are assumed to be partially exposed to a groundwater or soil environment.

The environments considered in the AMR are the environments that the TN-68 dry storage cask system SSCs and associated subcomponents normally experience.

Environmental stressors that are conditions not normally experienced (such as extreme cold), or that may be caused by a design or fabrication condition, are considered event-driven and are not aging-related. Such event-driven situations are evaluated and corrective actions, if any, implemented at the time of the event.

3.2.2 Identification of Aging Mechanisms and Aging Effects After the component material/environment combinations are identified, potential aging mechanisms are determined. NUREG-2214 [3-2] is reviewed to identify potential aging degradation mechanisms for different materials and environments.

Aging effects are the manifestation of aging mechanisms. In order to effectively manage an aging effect, it is necessary to determine the aging mechanisms that are potentially at work for a given material and environment combination. Therefore, the AMR process identifies both the aging effects and the associated aging mechanisms that cause them. Some aging mechanisms are only applicable at certain conditions such as high temperature or moisture. Each identified aging mechanism is characterized by a set of applicable conditions that must be met for the mechanism to occur. Given this evaluation process, each subcomponent that is subjected to AMR is evaluated to determine if the potential aging effects and mechanisms are credible considering the various material/environment combinations.

3.2.3 Determination of the Activities Required to Manage the Effects of Aging For each subcomponent with a credible aging mechanism and effect, a determination is made to ascertain whether the effect can be managed via a TLAA or if an AMP is necessary.

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CoC 1027 Renewal Application Revision 0 / April 2020 TLAAs are calculations or analyses used to demonstrate that an in-scope SSC will maintain its intended safety function throughout the PEO. TLAAs have a time-dependent operating life such as fatigue life (cycles), change in a mechanical property such as fracture toughness or strength of materials due to irradiation. The TLAAs associated with CoC 1027 are identified in Appendix 3A.

AMPs are developed for managing the effects of aging. As appropriate, an AMP is created to summarize the activities to monitor and manage the aging effects. The AMPs credited for managing the effects of aging degradation are presented in Chapter 4.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.3 Description of TN-68 Dry Storage Cask System Chapter 2 lists the following in-scope SSCs for the TN-68 dry storage cask system.

  • TN-68 cask
  • Spent fuel assemblies 3.3.1 Description of TN-68 Cask Subcomponents The following major in-scope subcomponents of the TN-68 dry storage cask are described in this section.
  • Cask body (and associated subcomponents)
  • Fuel basket
  • Cask lid
  • Cask seals Cask Body (and associated subcomponents)

The TN-68 cask containment vessel is comprised of an inner shell, which is a welded steel cylinder with an integrally-welded steel bottom; a welded flange forging (with stainless steel weld overlay); a bolted steel lid; and vent and drain port covers, cover seals and cover bolts. The confinement vessel is designed to the maximum practical extent as a Class I component in accordance with the provisions of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Subsection NB, with certain code exceptions as described in the UFSAR [3-3]. The inner shell is surrounded by a thick gamma shield. Radial neutron shielding around the gamma shield is provided by borated polyester resin encased in aluminum. A steel outer shell surrounds the radial neutron shielding. A relief valve at the top of the radial neutron shield provides a vent path for the buildup of gases.

Additional neutron shielding is provided by a polypropylene top neutron shield.

The cask external surfaces are painted for ease of decontamination. The cask sealing surfaces are clad with stainless steel.

Trunnions are provided at the cask upper and lower ends to permit cask movement and transport. A protective weather cover fits over the top of the cask. The cover is sealed with an elastomer O-ring.

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CoC 1027 Renewal Application Revision 0 / April 2020 Fuel Basket The fuel basket structure is made of an assembly of stainless steel cells joined by a fusion welding process to stainless steel plates. Above and below the plates are slotted borated aluminum or boron carbide/aluminum metal matrix composite plates (neutron poison plates) which form an egg-crate structure. The basket is supported laterally by aluminum rails which are attached to the periphery of the basket by welded studs. The fuel basket structure is designed to hold 68 fuel assemblies.

Cask Lid The cask lid is fabricated of steel and secured to the cask body with bolts. A steel shield plate is attached to the bottom of the lid to provide additional radiation shielding. The cask lid sealing arrangement consists of double metallic O-ring seals.

The cask lid and cask flange sealing surfaces are clad with a stainless steel weld overlay. Two penetrations are provided in the cask lid for cask venting and draining evolutions.

Cask Seals The cask lid, drain port, and vent port are equipped with Helicoflex double metallic O-ring seals. The metal seals consist of a stainless steel or nickel alloy liner, a nickel-based alloy spring, and an outer aluminum jacket.

3.3.2 Description of Storage Pad Subcomponents The storage pad is a reinforced concrete structure designed and constructed in accordance with codes and standards set by the general licensee. It is subject to site-specific foundation analyses and design considerations, including licensee-specific TN-68 cask loading configurations.

3.3.3 Description of Spent Fuel Assemblies Subcomponents The TN-68 dry storage cask system is designed to store 68 boiling water reactor (BWR) spent fuel assemblies with or without fuel channels. Table 3-1 summarizes the fuel types, cladding material, and burnup limits approved for storage in the TN-68 dry storage cask system.

Fuel Cladding and End Plugs The fuel rods consist of enriched UO2 pellets inserted into the cladding tubes. End plugs are seal welded to each end. The cladding and end plugs confine the fuel pellets and fission gases. Each rod is pressurized with helium during fabrication.

Fuel Channel The fuel channels of BWR fuel assemblies are mechanically attached and secured to the upper tie plate.

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CoC 1027 Renewal Application Revision 0 / April 2020 Spacer Grid Assemblies The grid assemblies provide support for the fuel rods, positioning them in a square array and maintaining the designed rod pitch.

Lower Tie Plate (and related subcomponents)

The lower tie plate functions as the bottom structural element of a fuel assembly and positions the fuel rods laterally.

Upper Tie Plate (and related subcomponents)

The upper tie plate functions as the top structural element of a fuel assembly. It also interfaces with the fuel assembly grapple as the lifting point for the fuel assembly.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4 Aging Management Review of Material/Environment 3.4.1 Materials Evaluated The materials of construction for the TN-68 dry storage cask system subcomponents, along with a material grouping, are listed in Table 3-5 through Table 3-8. The material groups represent a collection of individual material specifications that are susceptible to similar aging mechanism/effects and may, thus, be evaluated collectively. Table 3-2 provides a summary of the material groups.

3.4.2 Environments for the TN-68 Dry Storage Cask System SSCs The environments to which the TN-68 dry storage cask system subcomponents are exposed are listed in Table 3-5 through Table 3-8. These environments are those that are normally (continuously) experienced by the subcomponents while in storage, and are described below. Note that some of the subcomponents are exposed to two or more environments. In Table 3-5 through Table 3-8, the symbol (I) denotes environments seen by surfaces that are internal to the subcomponent, or surfaces that face towards the interior of the cask. The symbol (E) denotes environments seen by surfaces that are external to the subcomponent, or surfaces that face towards the exterior of the cask.

Internal to the TN-68 Cask Most of the internal subcomponents of the TN-68 cask are exposed to the inert gas (i.e., helium) environment inside the TN-68 cask cavity. The surfaces of some subcomponents are in contact with the surface of another subcomponent or material (i.e., an embedded-in-metal environment), e.g., the surfaces of the poison plate are in contact with the fuel compartment and structural plates. However, for the purpose of this AMR, if the space is not sealed, it is treated as though it is exposed to the helium environment. There are also other components that are in a fully encased environment, e.g., the exterior surface of the inner containment shell. The maximum fuel cladding temperature, cavity gas temperature, and cavity pressure during normal storage conditions were determined to be 649 °F, 422 °F, and 18.7 psig, respectively (Table 4B.3-1 and Table 4B.3-8 of Reference [3-3]). Table 3-3 summarizes the maximum component temperatures for the TN-68 cask during normal storage conditions.

The TN-68 cask internal components are exposed to significant neutron and gamma radiation.

External to the TN-68 Cask Each TN-68 cask is positioned for long-term storage on a reinforced concrete storage pad located outdoors; thus, the exterior surfaces of the TN-68 cask are exposed to all weather conditions, including insolation, wind, rain, snow, and site-specific ambient air conditions, including moist, possibly salt-laden atmospheric air, ambient temperatures, and humidity, i.e., an air-outdoor environment.

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CoC 1027 Renewal Application Revision 0 / April 2020 Storage pad The top surface of the storage pad is exposed to the air-outdoor environment. For the purpose of CoC renewal, no credit is taken for the cask sheltering the portion of the storage pad directly beneath the cask, i.e., it is assumed to be exposed to the air-outdoor environment. The below grade portion of the storage pad is exposed to a groundwater/soil environment.

Spent Fuel Assemblies Since the spent fuel assemblies are located within the sealed TN-68 cask, their subcomponents are exposed to the internal TN-68 cask environment, i.e., helium environment. During fabrication, each fuel pin is pressurized with helium. Hence, the internal environment of a fuel pin will consist of helium and fission gases, i.e., a helium environment.

3.4.3 Aging Effects Requiring Management This section evaluates known aging degradation mechanisms/effects for the TN-68 dry storage cask system materials of construction. For each mechanism, a determination is made of whether the mechanism is considered credible in each environment that the material is exposed to during the PEO. A credible aging mechanism is one that could manifest into an aging effect that affects an important-to-safety function during the PEO. The evaluation relied upon the information in NUREG-2214 [3-2] for identifying known aging degradation mechanisms and for determining if they are credible for the material/environment combination.

3.4.3.1 Aging Mechanism of Steel Material A review of Table 3-5 through Table 3-8 shows that the environments that the steel subcomponents are exposed to are:

  • Air-outdoor
  • Embedded-in-concrete
  • Fully encased The following aging mechanisms for steel material were evaluated to determine if they are credible in the environments that steel is exposed to:
  • General corrosion
  • Pitting and crevice corrosion
  • Galvanic corrosion
  • Microbiologically influenced corrosion (MIC)

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  • Creep
  • Fatigue
  • Thermal aging
  • Radiation embrittlement
  • Stress relaxation
  • Wear 3.4.3.1.1 General Corrosion of Steel Material General corrosion, also known as uniform corrosion, proceeds at approximately the same rate over a metal surface. Freely exposed steel surfaces in contact with moist air or water are subject to general corrosion.

In an air-outdoor environment, rain, fog, snow, and dew condensation can generate moisture layers on the steel surface that cause general corrosion.

In the embedded-in-concrete environment, the concrete provides an alkaline solution that passivates the steel. However, if the concrete degrades, the embedded steel could be exposed to water containing dissolved carbonates and chlorides and general corrosion is possible.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with steel will be limited. Similarly, there is very little moisture in a fully encased environment and, thus, the corrosion reaction with steel will be limited.

Therefore, general corrosion of the steel material is considered credible in air-outdoor or embedded-in-concrete environments, but not credible in helium or fully encased environments.

3.4.3.1.2 Pitting and Crevice Corrosion of Steel Material Pitting corrosion is a localized form of corrosion that is confined to a point or small area of a metal surface. It takes the form of cavities called pits. Crevice corrosion is another localized form of corrosion that occurs in a wetted environment when a crevice exists. It occurs more frequently in connections, lap joints, splice plates, bolt threads, under bolt heads, or at points of contact between metals and nonmetals.

Crevice corrosion is associated with stagnant or low-flow solutions. Steel is susceptible to pitting and crevice corrosion in an oxidizing and alkaline environment, especially in the presence of chlorides.

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CoC 1027 Renewal Application Revision 0 / April 2020 The potential to form aqueous electrolytes on the steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. These electrolytes could be conducive to pitting and crevice corrosion of steel. As such, there is the potential for pitting and crevice corrosion of steel subcomponents exposed to an air-outdoor environment.

In an embedded-in-concrete environment, if the concrete degrades with time, steel can be exposed to water containing dissolved carbonates and chlorides, which could be conducive to pitting and crevice corrosion as well.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with steel will be limited. Similarly, there is very little moisture in a fully encased environment and, thus, the corrosion reaction with steel will be limited.

Therefore, pitting and crevice corrosion of the steel material is considered credible in air-outdoor or embedded-in-concrete environments, but not credible in helium or fully encased environments.

3.4.3.1.3 Galvanic Corrosion of Steel Material Galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical contact in the presence of a conducting solution. Under these conditions, an electrolytic cell is formed, transmitting an electrical current between an anode (i.e.,

less noble material) and a cathode (i.e., more noble material). Oxidation occurs at the anode, and reduction occurs at the cathode. In a TN-68 cask, galvanic coupling exists between steel and other more noble materials such as stainless steel, graphite, and nickel.

The potential to form aqueous electrolytes on the steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. Because these electrolytes could initiate steel corrosion, and corrosion of steel is expected to be enhanced under galvanic coupling, loss of material due to galvanic corrosion of steel is considered credible in dissimilar metal couples.

In an embedded-in-concrete environment, the steel will not be in contact with a dissimilar metal or conductive material.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, very few aqueous electrolytes. Similarly, there is very little moisture in a fully encased environment and, thus, very few aqueous electrolytes.

Therefore, galvanic corrosion of the steel material is considered credible in an air-outdoor environment, but not credible in embedded-in-concrete, helium or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.1.4 Microbiologically Influenced Corrosion of Steel Material Microbiologically influenced corrosion is corrosion caused or promoted by the metabolic activity of microorganisms. Active microbial metabolism requires water in the form of water vapor, condensation, or deliquescence, and available nutrients to support microbial activity. Biofilms can form even under radiation environments. MIC is limited where relative humidity is below 90 percent and negligible for relative humidity below 60 percent. Although most of the evidence of MIC for metallic components is from conditions under which the metal surface is kept continuously wet, microorganisms can live in many environments, such as water, soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, sulfur/sulfide oxidizing bacteria, methane producers, organic acid-producing bacteria), fungi, and algae can develop.

The potential to form aqueous electrolytes on the steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or via deliquescence of deposited salts. These electrolytes have the potential to support microbial activity. However, there is no OE of MIC degradation of steel engineering components that are exposed to environments similar to those of dry storage cask systems, where continuous exposure to a relative humidity above 90 percent is not expected. The OE of MIC for metallic components is largely from instances in which the metal surface remained continuously wet. Because there is no applicable OE of MIC damage of steel under relevant atmospheric conditions, MIC is not considered credible in an air-outdoor environment.

In an embedded-in-concrete environment, if the concrete is exposed to a groundwater/soil environment and is degraded, the steel could be exposed to groundwater or soil. Under these conditions, the steel could be susceptible to MIC.

In the helium environment, there is very little residual water and nutrients in internal environments of a TN-68 cask following drying and refilling with inert helium gas.

Similarly, there is very little moisture and nutrients in a fully encased environment.

Therefore, MIC of the steel material is considered credible in an embedded-in-concrete environment if the concrete is exposed to a ground/soil environment, but is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.1.5 Stress Corrosion Cracking of Steel Material Stress corrosion cracking is the cracking of a metal produced by the combined action of corrosion and tensile stress (applied or residual). SCC is highly chemical-specific in that certain alloys are likely to undergo SCC only when exposed to a small number of chemical environments. SCC is the result of a combination of three factors: (1) a susceptible material, (2) exposure to a corrosive environment, and (3) tensile stresses.

High-strength steels with yield strengths greater than or equal to 150,000 pounds per square inch (150 ksi) have been found to be susceptible to SCC under exposure to aqueous electrolytes.

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CoC 1027 Renewal Application Revision 0 / April 2020 Since the steels used in the construction of the TN-68 dry storage cask system are not high-strength steels, i.e., they have a yield strength less than 150 ksi and, they are not susceptible to SCC.

Therefore, SCC of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.1.6 Creep of Steel Material Creep is the time-dependent, inelastic deformation that takes place at an elevated temperature and a constant stress. Because the deformation processes that produce creep are thermally activated, the rate of this time-dependent deformation is a strong function of the temperature. The creep rate also depends on the applied stress, but does not normally vary with the environment. As a general rule of thumb, at temperatures below 0.4Tm, where Tm is the melting point of the metal in Kelvin (K), thermal activation is insufficient to produce creep that could compromise the intended functions of SSCs. With a melting point of 1,789 °K (2,760 °F), temperatures of at least 716 °K (829 °F) are required to initiate creep in steels. However, the 0.4Tm rule of thumb underestimates the minimum creep temperature for steels, because it has been found that temperatures above 932 °F are required for creep in steels.

The highest temperatures within the TN-68 dry storage cask system are at locations close to the fuel rods. As shown in Table 3-3, the maximum fuel cladding temperature for fuel in a TN-68 cask was determined to be 649 °F at the beginning of storage.

Because the fuel rods are the only heat source within the system, they provide upper temperature limits for all subcomponents regardless of their environment. It is apparent from the temperatures in Table 3-3 that the steel subcomponents will not approach the minimum 932 °F required for creep to occur in steels.

Therefore, creep of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.1.7 Fatigue of Steel Material Fatigue is the progressive and localized structural damage that occurs when a material is subjected to cyclic loading. The only cyclic loading experienced by the TN-68 cask steel material is associated with thermal cycling. The only source of potential thermal fatigue of the TN-68 cask is ambient seasonal and daily temperature fluctuation. Due to the large thermal inertia of the TN-68 cask, it does not experience the full amplitude of ambient temperature cycles, and a gradual, long-term temperature decrease occurs during the course of storage.

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CoC 1027 Renewal Application Revision 0 / April 2020 Section 3.2.1.7 of NUREG-2214 [3-2] calls for a review of all fatigue analyses contained in the design basis documents to determine whether the renewal application adequately addresses the implications of extending the operating period to 60 years. It also says that if no fatigue analysis was performed in support of the component design, no action is required. The review of the design basis documents for fatigue analyses is in Appendix 3A. The review did not identify any fatigue analyses/evaluations for the TN-68 dry storage cask system components. Therefore, per NUREG-2214 [3-2], no action is required.

Therefore, fatigue of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.1.8 Thermal Aging of Steel Material The microstructures of most steels will change, given sufficient time at temperature, and this can affect mechanical properties. The degree to which thermal aging occurs depends on the steel grade and the exposure time and temperature.

The effects of elevated storage temperatures on material properties were evaluated during the initial license application. Carbon steels in the normalized condition (ferrite/pearlite microstructures) are commonly used in the petroleum and chemical industry with exposure temperatures of approximately 400 °C (752 °F). Per Table 3-3 the maximum temperature of a steel subcomponent (i.e., the inner shell) is only 340 °F at the beginning of the initial storage period. Therefore, it can be concluded that thermal aging is not expected to produce degradation of the mechanical properties of steels in the PEO.

Therefore, thermal aging of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.1.9 Radiation Embrittlement of Steel Material Embrittlement of metals may occur under exposure to neutron radiation. Depending on the neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, reduced fracture toughness, and decreased resistance to cracking.

Section 3.2.1.9 of NUREG-2214 [3-2] states that neutron fluence levels greater than 1019 neutrons per square centimeter (n/cm2) are required to produce a measureable degradation of the mechanical properties. NUREG-2214 [3-2] then describes a bounding calculation that shows estimated fluence level within a dry storage cask system is three orders of magnitude below the levels reported to degrade the fracture resistance of carbon and alloy steels.

Therefore, radiation embrittlement of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.1.10 Stress Relaxation of Steel Material Stress relaxation of bolting or other tightening subcomponents is the steady loss of elastic stress in a loaded part due to atomic movement at elevated temperature. It results in a loss of clamping forces or preload in a heavily loaded joint. Stress relaxation is a strong function of temperature and bolt material. It also depends on geometry of the bolt and thread quality. It decreases with time, as the tensile stress in the bolt decreases.

Research has demonstrated that the residual stress of carbon steel bolts due to relaxation is about 85 percent of the initial applied stress at temperatures greater than about 212 °F. Table 3-3 shows that the initial temperature of the lid is greater than the 212 °F threshold. Therefore, bolts in contact with the lid (i.e., in an air-outdoor environment) may also be above the stress relaxation temperature threshold. A review of Table 3-5 through Table 3-8 shows that there are no bolts in an embedded-in-concrete, helium, or fully encased environment.

Therefore, stress relaxation of the steel bolt material is considered credible in an air-outdoor environment, but not credible in embedded-in-concrete, helium, or fully encased environments.

3.4.3.1.11 Wear of Steel Material Rolling contact wear results from the repeated mechanical stressing of the surface of a body rolling on another body.

There are no steel bodies rolling on another body while the TN-68 casks are in storage.

Therefore, wear of the steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.2 Aging Mechanism of Stainless Steel Material A review of Table 3-5 through Table 3-8 shows that the environments that the stainless steel subcomponents are exposed to are:

  • Air-outdoor
  • Fully encased The following aging mechanisms for stainless steel material were evaluated to determine if they are credible in the environments that stainless steel is exposed to:
  • General corrosion
  • Pitting and crevice corrosion
  • Galvanic corrosion Page 3-16

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  • Creep
  • Fatigue
  • Thermal aging
  • Radiation embrittlement
  • Stress relaxation
  • Wear 3.4.3.2.1 General Corrosion of Stainless Steel Material Stainless steels exhibit passive behavior in all environments, resulting in negligible general corrosion rates.

Therefore, general corrosion of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.2.2 Pitting and Crevice Corrosion of Stainless Steel Material Pitting corrosion is a localized form of corrosion that is confined to a point or small area of a metal surface. It takes the form of cavities called pits. Crevice corrosion is another localized form of corrosion that occurs in a wetted environment when a crevice exists that allows a corrosive environment to develop. It occurs more frequently in connections, lap joints, splice plates, bolt threads, under bolt heads, or at points of contact between metals and nonmetals. Crevice corrosion is associated with stagnant or low-flow solutions. Stainless steel is known to be susceptible to pitting and crevice corrosion.

The potential to form aqueous electrolytes on the steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. These electrolytes could be conducive to pitting and crevice corrosion of stainless steel. As such, there is the potential for pitting and crevice corrosion of stainless steel subcomponents exposed to an air-outdoor environment.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with stainless steel will be limited. Similarly, there is very little moisture in a fully encased environment and, thus, the corrosion reaction with stainless steel will be limited.

Therefore, pitting and crevice corrosion of the stainless steel material is considered credible in air-outdoor environment, but not credible in helium, or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.2.3 Galvanic Corrosion of Stainless Steel Material Galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical contact in the presence of a conducting solution. Under these conditions, an electrolytic cell is formed, transmitting an electrical current between an anode (i.e.,

less noble material) and a cathode (i.e., more noble material). Oxidation occurs at the anode, and reduction occurs at the cathode. Galvanic coupling exists between stainless steel and other more noble materials such as graphite.

The potential to form aqueous electrolytes on the stainless steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. Because electrolytes conducive to galvanic corrosion may exist, galvanic corrosion of stainless steel in contact with graphite lubricants is considered credible.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, very few aqueous electrolytes. Similarly, there is very little moisture in a fully encased environment and, thus, very few aqueous electrolytes.

Therefore, galvanic corrosion of the stainless steel material is considered credible in an air-outdoor environment when in contact with a graphite lubricant, but not credible in helium, or fully encased environments.

3.4.3.2.4 Microbiologically Influenced Corrosion of Stainless Steel Material As discussed in Section 3.4.3.1.4, MIC is caused or promoted by the metabolic activity of microorganisms. Microorganisms can live in many environments, such as water, soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, sulfur/sulfide oxidizing bacteria, methane producers, and organic acid-producing bacteria), fungi, and algae can develop.

The potential to form aqueous electrolytes on the stainless steel surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. These electrolytes have the potential to support microbial activity. However, there has not yet been any OE of MIC in atmospheric environments where stainless steel surfaces are only intermittently wetted. Because there is no applicable OE of MIC damage of stainless steel under relevant atmospheric conditions, MIC is not considered credible in an air-outdoor environment.

In the helium environment, there is very little residual water and nutrients in internal environments of a TN-68 cask following drying and refilling with inert helium gas.

Similarly, there is very little moisture and nutrients in a fully encased environment.

Therefore, MIC of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.2.5 Stress Corrosion Cracking of Stainless Steel Material Stress corrosion cracking is the cracking of a metal produced by the combined action of corrosion and tensile stress and is highly chemical-specific. Most ferritic and duplex stainless steels are either immune or highly resistant to SCC; however, all austenitic grades, especially Types 304, 304L, 304LN, 316, 316L, and 316LN, are susceptible to chloride-induced SCC in the normal wrought condition. This susceptibility increases when the material is sensitized. In the welded condition, the heat affected zone (HAZ),

which is a thin band located adjacent to the weld, can be sensitized by the precipitation of carbides that extract chromium out of the metal matrix.

In the helium environment, there is a lack of halides and very little residual water in internal environments of a TN-68 following drying and refilling with inert helium gas and, thus, SCC is not considered credible. Similarly, there is a lack of halides and very little moisture in a fully encased environment and, thus, SCC is not considered credible.

The construction of the TN-68 dry storage cask system does not involve welding of stainless steel material together in an air-outdoor environment.

Therefore, SCC of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.2.6 Creep of Stainless Steel Material As discussed in Section 3.4.3.1.6, as a general rule of thumb, at temperatures below 0.4Tm, where Tm is the melting point of the metal in Kelvin, thermal activation is insufficient to produce creep that could compromise the intended functions of SSCs.

The term stainless steel covers a wide range of compositions and microstructures, including austenitic, ferritic, martensitic, duplex, and precipitation hardening stainless steels. Using the melting point temperature for austenitic or 300 series stainless steels (because they are most commonly used in the dry storage cask systems and have the lowest melting point) of 1,698 °K (2,597 °F), temperatures of at least 679 °K (763 °F) are required to initiate creep in the austenitic stainless steels.

The highest temperatures within the TN-68 dry storage cask system are at locations close to the fuel rods. As shown in Table 3-3, the maximum fuel cladding temperature for fuel in a TN-68 cask was determined to be 649 °F at the beginning of storage.

Because the fuel rods are the only heat source within the system, they provide upper temperature limits for all subcomponents regardless of their environment. It is apparent from the temperatures in Table 3-3 that internal stainless steel subcomponents will not approach the minimum 763 °F temperature that has been found to be required for creep to occur in stainless steels.

Therefore, creep of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.2.7 Fatigue of Stainless Steel Material Fatigue is the progressive and localized structural damage that occurs when a material is subjected to cyclic loading. The only cyclic loading experienced by the TN-68 cask stainless steel material is associated with thermal cycling. The only source of potential thermal fatigue of the TN-68 cask is ambient seasonal and daily temperature fluctuation. Due to the large thermal inertia of the TN-68 cask, it does not experience the full amplitude of ambient temperature cycles, and a gradual, long-term temperature decrease occurs during the course of storage.

Section 3.2.1.7 of NUREG-2214 [3-2] calls for a review of all fatigue analyses contained in the design basis documents to determine whether the renewal application adequately addresses the implications of extending the operating period to 60 years. It also says that if no fatigue analysis was performed in support of the component design, no action is required. The review of the design basis documents for fatigue analyses is in Appendix 3A. The review did not identify any fatigue analyses/evaluations for the TN-68 dry storage cask system components. Therefore, per NUREG-2214 [3-2], no action is required.

Therefore, fatigue of the stainless steel material is not credible in air-outdoor, embedded-in-concrete, helium, or fully encased environments.

3.4.3.2.8 Thermal Aging of Stainless Steel Material The microstructures of most stainless steels will change, given sufficient time at temperature, and these changes may alter the materials strength and fracture toughness. For stainless steel subcomponents, the thermal aging process differs for welded and non-welded subcomponents.

Welded Austenitic Stainless Steel Subcomponents in Helium The ferrite present in austenitic stainless steel welds can transform by spinodal decomposition to form iron-rich alpha and chromium-rich alpha prime phases, and further aging can produce an intermetallic G-phase. The spinodal decomposition and the formation of the intermetallic G-phase takes place during extended exposure to temperatures between 300 and 400 °C (572 and 752 °F). The highest temperatures within the TN-68 dry storage cask system are at locations close to the fuel rods. As shown in Table 3-3, the maximum fuel cladding temperature for fuel in a TN-68 cask was determined to be 649 °F at the beginning of storage. Because the fuel rods are the only heat source within the system, they provide upper temperature limits for all subcomponents regardless of their environment. It is apparent from the temperatures in Table 3-3 that internal stainless steel subcomponents may be above the 300 °C (572

°F) minimum temperature required for these phase changes.

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CoC 1027 Renewal Application Revision 0 / April 2020 Based on Charpy impact toughness testing of cast duplex stainless steels, it is concluded that ferrite levels above 15 percent are required for significant embrittlement, because ferrite resides in discrete islands below this level and does not provide a continuous low-toughness fracture path. Because most welds contain around 4 to 15 percent ferrite, substantial embrittlement of austenitic stainless steel welds is not expected. NUREG/CR-6428 [3-7] concluded that thermal aging produced moderate decreases (no more than 25 percent) in the upper shelf Charpy impact energy and relatively small decreases in the fracture toughness of a wide range of austenitic welds. Although the phase changes associated with thermal embrittlement of austenitic stainless steel welds could take place in subcomponents near the fuel within the 60-year timeframe, the minor reductions in fracture toughness that would be produced in the weld indicate that this is not a credible aging mechanism for subcomponents in proximity to the fuel rods.

Non-Welded Austenitic Stainless Steel Subcomponents in Helium Because the phase changes described previously occur only within the ferrite-containing, HAZ of a weld, embrittlement will not occur in austenitic stainless steel subcomponents that do not contain a weld.

Stainless Steel Subcomponents in Air-Outdoor, and Fully Encased Environments Because the peak temperatures for stainless steel subcomponents exposed to air-outdoor, and fully encased environments are below the temperature required for the phase changes associated with thermal embrittlement of stainless steels, thermal aging is not considered credible for these subcomponents.

Summary Therefore, thermal aging of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.2.9 Radiation Embrittlement of Stainless Steel Material Embrittlement of metals may occur under exposure to neutron radiation. Depending on the neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, reduced fracture toughness, and decreased resistance to cracking.

Section 3.2.2.9 of NUREG-2214 [3-2] states that neutron fluence levels greater than 1020 n/cm2 are required to produce a measureable degradation of the mechanical properties. NUREG-2214 [3-2] then describes a bounding calculation that showed estimated fluence level within a dry storage cask system is four orders of magnitude below the levels that would degrade the mechanical properties of stainless steel.

Therefore, radiation embrittlement of the stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.2.10 Stress Relaxation of Stainless Steel Material Stress relaxation of bolting or other tightening subcomponents is the steady loss of elastic stress in a loaded part due to atomic movement at elevated temperature. It results in a loss of clamping forces or preload in a heavily loaded joint. Stress relaxation is a strong function of temperature and bolt material. It also depends on geometry of the bolt and thread quality. It decreases with time, as the tensile stress in the bolt decreases.

Research has demonstrated that the loss of initial applied stress in austenitic stainless steel bolting due to stress relaxation is negligible at temperatures below 572 °F.

Table 3-3 shows that the initial maximum temperature of the lid is less than this threshold. Therefore, stainless steel bolts in contact with the lid (i.e., in an air-outdoor environment) are below the stress relaxation temperature threshold. A review of Table 3-5 through Table 3-8 shows that there are no bolts in a helium or fully encased environment.

Therefore, stress relaxation of the stainless steel bolt material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.2.11 Wear of Stainless Steel Material Rolling contact wear results from the repeated mechanical stressing of the surface of a body rolling on another body.

There are no stainless steel bodies rolling on another body while TN-68s are in storage. While the upper trunnions are bolted to the TN-68 cask, Section 1.2.1 of the UFSAR [3-3] states that they are not intended to be removed during storage and, thus, are not intended to be reused.

Therefore, wear of other stainless steel material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.3 Aging Mechanisms of Aluminum Material A review of Table 3-5 through Table 3-8 shows that the only environment that the aluminum subcomponents are exposed to is:

  • Fully encased
  • Air-outdoor The following aging mechanisms for aluminum material were evaluated to determine if they are credible in the environments that aluminum is exposed to:
  • General corrosion
  • Pitting and crevice corrosion Page 3-22

CoC 1027 Renewal Application Revision 0 / April 2020

  • Galvanic corrosion
  • Creep
  • Fatigue
  • Thermal aging
  • Radiation embrittlement 3.4.3.3.1 General Corrosion of Aluminum Material General corrosion, also known as uniform corrosion, proceeds at approximately the same rate over a metal surface. Freely exposed aluminum surfaces in contact with moist air or water are subject to general corrosion. The corrosion rate depends on solution composition, pH, and temperature. The corrosion rate of aluminum is normally controlled by the formation of a passive film of Al2O3 at the metal and water interface.

In an air-outdoor environment, rain, fog, snow, and dew condensation can generate moisture layers on the aluminum surface that cause general corrosion.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with aluminum will be limited. Similarly, there is very little moisture in a fully encased environment and, thus, the corrosion reaction with aluminum will be limited.

Therefore, general corrosion of the aluminum material is credible in an air-outdoor environment but not credible in helium, or fully encased environments.

3.4.3.3.2 Pitting and Crevice Corrosion of Aluminum Material As discussed in Section 3.4.3.1.2, pitting corrosion is a localized form of corrosion that is confined to a point or small area of a metal surface, and crevice corrosion occurs in a wetted environment when a crevice exists that allows a corrosive environment to develop in a component. Aluminum and its alloys form a passive film on the surface. However, localized corrosion in the form of pitting or crevice corrosion could occur on aluminum subcomponents, especially in the presence of halides.

In an air-outdoor environment, rain, fog, snow, and dew condensation can generate moisture layers on the aluminum surface that cause corrosion reaction.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with aluminum will be limited. Similarly, there is very little moisture in a fully encased environment and, thus, the corrosion reaction with aluminum will be limited.

Page 3-23

CoC 1027 Renewal Application Revision 0 / April 2020 Therefore, pitting and crevice corrosion of the aluminum material is credible in an air-outdoor environment but not credible in helium, or fully encased environments.

3.4.3.3.3 Galvanic Corrosion of Aluminum Material As discussed in Section 3.4.3.1.3, galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical contact in the presence of a conducting solution. In dry storage cask systems, galvanic coupling exists between aluminum and steel, and stainless steel (where aluminum is less noble in each case).

The potential to form aqueous electrolytes on surfaces exposed to an air-outdoor environment is present via direct exposure to precipitation or through deliquescence of deposited salts. Because these electrolytes could initiate aluminum corrosion, and corrosion of aluminum is expected to be enhanced under galvanic coupling, loss of material due to galvanic corrosion of aluminum is considered credible in dissimilar metal couples.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, very few aqueous electrolytes. Similarly, there is very little moisture in a fully encased environment and, thus, very few aqueous electrolytes.

Therefore, galvanic corrosion of the aluminum material is considered credible in an air-outdoor environment, but not credible in helium or fully encased environments.

3.4.3.3.4 Microbiologically Influenced Corrosion of Aluminum Material Microbiologically influenced corrosion (MIC) is corrosion caused or promoted by the metabolic activity of microorganisms. Active microbial metabolism requires water in the form of water vapor, condensation, or deliquescence, and available nutrients to support microbial activity. Biofilms can form even under radiation environments.

Microbiologically influenced corrosion is limited where relative humidity is below 90 percent and negligible for relative humidity below 60 percent. Although most of the evidence of MIC for metallic components is from conditions under which the metal surface is kept continuously wet, microorganisms can live in many environments, such as water, soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, sulfur/sulfide oxidizing bacteria, methane producers, and organic acid-producing bacteria), fungi, and algae can develop.

However, there is no operating experience of MIC degradation of aluminum engineering components that operate in environments similar to those of dry cask storage systems. All of the OE of MIC for metallic components is from conditions in which the metal surface is kept continuously wet. Due to the absence of any applicable experience of MIC damage of aluminum components under atmospheric conditions, MIC is not considered to be significant in an air-outdoor environment.

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CoC 1027 Renewal Application Revision 0 / April 2020 In the helium environment, there is very little residual water and nutrients in internal environments of a TN-68 cask following drying and refilling with inert helium gas.

Similarly, there is very little moisture and nutrients in a fully encased environment.

Therefore, MIC of the aluminum material is not credible in air-out door, helium or fully encased environments.

3.4.3.3.5 Creep of Aluminum Material Creep is the time-dependent, inelastic deformation that takes place at an elevated temperature and a constant stress. Because the deformation processes that produce creep are thermally activated, the rate of this time-dependent deformation is a strong function of the temperature. The creep rate also depends on the applied stress but does not generally vary with the environment. As a general rule of thumb, at temperatures below 0.4Tm, where Tm is the melting point of the metal in Kelvin, thermal activation is insufficient to produce creep that could compromise the intended functions of SSCs.

With a melting point of 911 to 930 °K (1,180 to 1,215 °F), temperatures of at least 364 to 372 °K (196 to 210 °F) are required to initiate creep in aluminum. These temperatures are consistent with other research that which indicates that creep in aluminum is possible at temperatures greater than 212 °F. Microstructure also plays a significant role in the resistance to creep of a metal. Hence, while this 212 °F minimum temperature for creep is representative for pure aluminum, creep in precipitation-hardened aluminum alloys does not become significant until about 392 °F. Additionally, at temperatures near these threshold values, high stresses are required to produce creep.

Subcomponents that do not serve a structural function are not under loads other than their own weight, and in many instances, their weight is also supported by adjacent structures. Due to the minimal applied loads, creep of nonstructural subcomponents will not produce significant damage during the 60-year timeframe.

A review of Table 3-5 through Table 3-8 shows that the only aluminum sub-components with a structural function are the basket rails. These basket rails are constructed out of precipitation-hardened 6061-T6 aluminum. Table 3-3 shows that the maximum temperature for the basket rails is 402 °F. While this value is slightly above the creep threshold temperature of 392 °F for precipitation-hardened aluminum alloys, it is expected to drop below the threshold value by the end of the initial storage period. In addition, the basket rails are under no loads other than their own weight and, thus, are not expected to experience creep.

Therefore, creep of the aluminum material is not credible in air-outdoor, helium or fully encased environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.3.6 Fatigue of Aluminum Material Fatigue is the progressive and localized structural damage that occurs when a material is subjected to cyclic loading. The only cyclic loading experienced by the TN-68 cask aluminum material is associated with thermal cycling. The only source of potential thermal fatigue of the TN-68 cask is ambient seasonal and daily temperature fluctuation. The TN-68 cask does not experience the full amplitude of ambient temperature cycles, and a gradual, long-term temperature decrease occurs during the course of storage. The seasonal and daily variations in ambient conditions are ameliorated by the thermal mass of the TN-68 cask.

Section 3.2.3.6 of NUREG-2214 [3-2] calls for a review of all fatigue analyses contained in the design basis documents to determine whether the renewal application adequately addresses the implications of extending the operating period to 60 years. It also says that if no fatigue analysis was performed in support of the component design, no action is required. The review of the design basis documents for fatigue analyses is in Appendix 3A. The review did not identify any fatigue analyses/evaluations for the TN-68 dry storage cask system components. Therefore, per NUREG-2214 [3-2], no action is required.

Therefore, fatigue of the aluminum material is not credible in air-outdoor, helium, or fully encased environments.

3.4.3.3.7 Thermal Aging of Aluminum Material The microstructures of most aluminum alloys will change, given sufficient time at temperature, and this can affect mechanical properties. The effect of thermal aging (i.e., loss of strength) will depend on the time at temperature and the microstructure and chemical composition of the aluminum subcomponents. This loss of strength could be an issue for any subcomponents that perform a structural function.

A review of Table 3-5 through Table 3-8 shows that the only aluminum sub-components that have a structural function are the basket rails. These subcomponents are constructed out of precipitation-hardened 6061-T6 aluminum.

Research has shown that when alloy 6061-T6 is held at 392 °F its yield strength drops from approximately 18 ksi at 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (1.14 years) to approximately 11.5 ksi at 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (11.4 years). Because of this sensitivity to exposure time, ASME B&PV Code Section II requires that time-dependent properties be used for exposures above 350 °F for this alloy. UFSAR [3-3] Section 3.3.2 states that the material properties of the Aluminum alloy (6061-T6) were taken from the ASME Code.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-3 shows that when the cask is initially placed in storage, the maximum temperature for the basket rails is 402 °F. As the temperatures inside the cask decrease during the initial storage period (e.g., Section 3.2.3.7 of NUREG-2214 [3-2] describes a 242 °F decrease in the maximum fuel cladding temperature) it is expected that the basket rails temperatures will drop below 350 °F by the end of the initial storage period. Therefore, thermal aging of aluminum is not expected to be a concern during the PEO. In addition, the summary of the basket rail structural analysis in UFSAR [3-3] Section 3B.6 shows that the limiting stress in the basket aluminum rail meets the allowable stress limit for a 98 g side drop load which is much higher than the maximum calculated g-load of 77 g. Note that these g-loadings are at the top of the basket while the area of the higher temperature is in the middle of the basket where the g-loads are much lower. Due to high margin of safety compared against the ASME code allowable limit, it is reasonable to conclude that any prolonged elevated temperature exposure resulting in reduced strength of aluminum alloy will not affect the TN-68 basket rails. Therefore, any thermal aging of the basket rails is not expected to result in the loss of their structural intended safety function during the PEO.

Therefore, thermal aging of the aluminum material is not credible in air-outdoor, helium or fully encased environments.

3.4.3.3.8 Radiation Embrittlement of Aluminum Material Embrittlement of metals may occur under exposure to neutron radiation. Depending on the neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, reduced fracture toughness, and decreased resistance to cracking.

Section 3.2.3.8 of NUREG-2214 [3-2] describes data showing that the mechanical properties of aluminum are not degraded at neutron fluence levels on the order of 1020 n/cm2. Section 3.2.1.9 of NUREG-2214 [3-2] describes a bounding calculation that showed estimated fluence level within a dry storage cask system is four orders of magnitude below the levels reported to degrade the mechanical properties of aluminum.

Therefore, radiation embrittlement of the aluminum material is not credible in air-out door, helium, and fully encased environments.

3.4.3.4 Aging Mechanism of Nickel Alloy Material A review of Table 3-5 through Table 3-8 shows that the only subcomponent with material in the nickel alloy material group is a portion of the metallic O-rings (i.e., the liner) exposed to:

  • Helium The following aging mechanisms for nickel alloy material were evaluated to determine if they are credible in the environment that the O-ring liner, i.e., nickel alloy material, is exposed to:
  • General corrosion Page 3-27

CoC 1027 Renewal Application Revision 0 / April 2020

  • Pitting and crevice corrosion
  • Fatigue
  • Radiation embrittlement
  • Wear 3.4.3.4.1 General Corrosion of Nickel Alloy Material There is very little residual water in the interspace of the metallic O-rings and in the over pressure monitoring gas.

Therefore, general corrosion of the nickel alloy material is not credible in a helium environment.

3.4.3.4.2 Pitting and Crevice Corrosion of Nickel Alloy Material Pitting and crevice corrosion require the presence of aqueous electrolytes to form a corrosive environment. There is very little residual water in the interspace of the metallic O-rings and in the over pressure monitoring gas to generate aqueous electrolytes.

Therefore, pitting and crevice corrosion of the nickel alloy material is not credible in a helium environment.

3.4.3.4.3 Microbiologically Influenced Corrosion of Nickel Alloy Material Active microbial metabolism requires water in the form of water vapor, condensation, or deliquescence, and available nutrients to support microbial activity. There is very little residual water in the interspace of the metallic O-rings and in the over pressure monitoring gas to support MIC.

Therefore, MIC of the nickel alloy material is not credible in a helium environment.

3.4.3.4.4 Stress Corrosion Cracking of Nickel Alloy Material SCC is the cracking of a metal produced by the combined action of corrosion and tensile stress and is highly chemical-specific. There is very little residual water in the interspace of the metallic O-rings and in the over pressure monitoring gas to support a corrosive environment. In addition, the metallic seals are crushed during installation and, as such, there are no tensile stresses.

Therefore, SCC of the nickel alloy material is not credible in a helium environment.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.4.5 Fatigue of Nickel Alloy Material Fatigue is the progressive and localized structural damage that occurs when a material is subjected to cyclic loading. The only cyclic loading experienced by the TN-68 cask nickel alloy material is associated with thermal cycling due to ambient seasonal and daily temperature fluctuation. However, the metallic seals are crushed during installation and thermal cycling of the liner will not affect the ability of the aluminum jacket to fulfill the confinement function.

Therefore, fatigue of the nickel alloy material is not credible in a helium environment.

3.4.3.4.6 Radiation Embrittlement of Nickel Alloy Material Embrittlement of metals may occur under exposure to neutron radiation. Depending on the neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, reduced fracture toughness, and decreased resistance to cracking. The function of the seal nickel alloy liner and spring is to maintain the shape of the seal prior to installation. During installation, the seals are crushed including the liner and spring. Therefore, radiation embrittlement of the seal liner and spring will not affect the ability of the aluminum jacket to fulfill the confinement function.

Therefore, radiation embrittlement of the nickel alloy material is not credible in a helium environment.

3.4.3.4.7 Wear of Nickel Alloy Material Rolling contact wear results from the repeated mechanical stressing of the surface of a body rolling on another body.

There are no nickel alloy bodies rolling on another body while TN-68 casks are in storage.

Therefore, wear of the nickel alloy material is not credible in a helium environment.

3.4.3.5 Aging Mechanism of Polymer Material A review of Table 3-5 through Table 3-8 shows that the environment that the polymer subcomponents are exposed to is:

  • Fully Encased The following aging mechanisms for polymer material were evaluated to determine if they are credible in the environment that polymer material is exposed to:
  • Thermal aging
  • Radiation embrittlement Page 3-29

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.5.1 Boron Depletion of Polymer Material Table 5.1-1 of the UFSAR [3-3], lists the thickness of the radial neutron shield as 6 inches (15.24 cm) and Table 5.3-5 of the UFSAR [3-3] list the atom fractions of the materials used in the shielding models for the radial neutron shield along with the density of the radial neutron shield. The areal atom density of each element may be determined using the density of the neutron shield, the atom fractions, the molar mass and the thickness of the shield as follows:

  • The relative mass of each element is determined by multiplying the atom fraction and the molar mass of the element then dividing by Avogadros number (6.0221*1023) particles per mole.
  • The weight fraction of each element is determined by dividing the relative mass by the sum of all relative masses.
  • The density of the element is determined by multiplying the weight fraction and the density of the radial neutron shield.
  • Finally the areal density is determined by multiplying the density, Avogadros number and the thickness of the radial neutron shield then dividing by the molar mass.

Relative Areal Atom Molar Mass Weight Element Mass Density Density Fraction (g/mole) Fraction (grams) (atoms/cm2) 27 Al 0.10331 26.982 4.629 2.851x10-1 4.810x10-1 1.636x1023 12 C 0.24658 12.000 4.914 3.027x10-1 5.106x10-1 3.905x1023 16 O 0.21985 15.995 5.839 3.596x10-1 6.066x10-1 3.480x1023 1

H 0.42207 1.008 7.065x10-1 4.352x10-3 7.342x10-3 6.555x1023 10 B 0.00164 10.013 2.727x10-2 1.680x10-3 2.834x10-3 2.598x1021 11 B 0.00655 11.009 1.197x10-1 7.373x10-3 1.244x10-2 1.037x1022 Total 16.235 The areal density of boron-10 (2.598x1021 atom/cm2) is on the order of the boron-10 areal density used in the evaluation of boron depletion performed in Section 3.4.1.1 of NUREG-2214 [3-2]. That section of NUREG-2214 concluded that with a typical neutron flux within a dry storage cask system, boron depletion would only be 0.0002 percent of the available boron-10 atoms.

Therefore, boron depletion of the polymer material is not credible in a fully encased environment.

Page 3-30

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.5.2 Thermal Aging of Polymer Material Polymers may be susceptible to heat-induced changes to material properties and configuration due to a number of mechanisms. At elevated temperatures, the long chain backbone of a polymer can undergo molecular scission (breaking) and cross linking. Also, gaseous products may be formed, including H2, CH4, and CO2. These reactions may cause embrittlement, shrinkage/cracking, decomposition, and changes in physical configuration (e.g., loss of hydrogen or water).

In the TN-68 casks, the polymer materials are fully encased and have no structural integrity function. Therefore, embrittlement of the material is not an aging effect that needs to be managed.

Appendix 9A of the UFSAR [3-3], describes the thermal stability of the TN-68 cask radial neutron shield material. Interpolation of test data indicates an exponential weight loss that rapidly approaches a maximum weight loss value of 1.3%. This maximum weight loss, i.e., off gassing or change in physical configuration, would occur during the initial license period and, thus, not be affected by the PEO.

Therefore, decomposition, changes in physical configuration, and the buildup of flammable gas due to thermal aging in the polymers are not aging effects that must be managed during the PEO.

Shrinkage/cracking due to thermal aging could locally displace shielding material and potentially diminish shielding effectiveness. Although this effect would most likely occur during the initial license period, it should be managed during the PEO.

Therefore, thermal aging of the polymer material is considered credible in a fully encased environment.

3.4.3.5.3 Radiation Embrittlement of Polymer Material Radiation can alter polymer structures by molecular scission and cross-linking to reduce ductility, fracture toughness, and resistance to cracking.

The aging management review of radiation embrittlement of the polymer material is the same as for thermal aging of the polymer material. Therefore, embrittlement of the material, decomposition, changes in physical configuration, and the buildup of flammable gas due to radiation of the polymers are not aging effects that must be managed during the PEO. However, shrinkage/cracking is an aging effect that should be managed during the PEOs.

Therefore, radiation embrittlement of the polymer material is considered credible in a fully encased environment.

3.4.3.6 Aging Mechanism of Borated Aluminum Material A review of Table 3-5 through Table 3-8 shows that the environment that the borated aluminum subcomponents are exposed to is:

Page 3-31

CoC 1027 Renewal Application Revision 0 / April 2020

  • Helium The following aging mechanisms for borated aluminum material were evaluated to determine if they are credible in the environment that borated aluminum is exposed to:
  • General corrosion
  • Galvanic corrosion
  • Creep
  • Thermal aging
  • Radiation embrittlement
  • Wet corrosion and blistering 3.4.3.6.1 General Corrosion of Borated Aluminum Material Because aluminum is present as a continuous matrix (borated aluminum and aluminum metal-matrix composites) or used as an outer cladding (Boral), the degree of general corrosion is considered to be largely governed by the corrosion of aluminum.

As discussed in Section 3.4.3.3.1, in the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, the corrosion reaction with aluminum will be limited.

Therefore, general corrosion of the borated aluminum material is not credible in a helium environment.

3.4.3.6.2 Galvanic Corrosion of Borated Aluminum Material As discussed in Section 3.4.3.1.3, galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical contact in the presence of a conducting solution. The aluminum-based neutron poison materials used inside dry storage cask systems can be in galvanic contact with the stainless steel fuel compartments, where aluminum is less noble.

In the helium environment, there is very little residual water in internal environments of a TN-68 cask following drying and refilling with inert helium gas and, thus, very few aqueous electrolytes.

Therefore, galvanic corrosion of the borated aluminum material is not credible in a helium environment.

Page 3-32

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.6.3 Boron Depletion of Borated Aluminum Material Boron depletion refers to the loss of the capability of a material to absorb neutrons when the neutron fluence significantly consumes boron-10 atoms. At typical levels of neutron flux and boron-10 concentration, the neutron dose after 60 years would deplete at most 0.0002 percent of the available boron-10 atoms. Using the highest expected neutron flux and the lowest boron-10 concentration as a worst-case scenario, only 0.02 percent of the available boron-10 atoms would be depleted after 60 years, which is too small to challenge the criticality control function of the neutron poisons.

Therefore, boron depletion of the borated aluminum material is not credible in a helium environment.

3.4.3.6.4 Creep of Borated Aluminum Material Creep is the time-dependent inelastic deformation that takes place at an elevated temperature and a constant stress. Because the deformation processes that produce creep are thermally-activated, the rate of this time-dependent deformation is a strong function of the temperature. Because aluminum is present as a continuous matrix or as an external cladding in the neutron poison plates, and aluminum has a lower melting point than the other portions of the material microstructures, the creep behavior of poison materials is considered to be governed by the behavior of aluminum.

As discussed in Section 3.4.3.3.5, as a general rule of thumb, at temperatures below 0.4Tm, where Tm is the melting point of the metal in Kelvin, thermal activation is insufficient to produce creep that could compromise the intended functions of SSCs.

Applying the 0.4Tm rule, the critical creep temperature for aluminum is 212 °F. As shown in Table 3-3, the maximum basket temperature was determined to be 624 °F at the beginning of storage. It is apparent from this temperature that the borated aluminum plates could be exposed to temperatures above the minimum creep temperatures for aluminum.

Because temperatures within the TN-68 cask have the potential to exceed the minimum creep temperature of aluminum, it is necessary to consider the load applied to the subcomponent to determine whether significant creep deformation will occur, as well as the specific application, to determine whether the creep affects safety. As shown in Table 3-5 through Table 3-8, the borated aluminum subcomponents do not serve a structural function and are not expected to be under loads other than their own weight. In addition, the borated aluminum plates are supported by adjacent subcomponents. Because of the minimal applied loads and presence of adjacent supporting structures, the impact of creep on the intended safety functions function of the borated aluminum subcomponents is not considered credible.

Therefore, creep of the borated aluminum material is not credible in a helium environment.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.6.5 Thermal Aging of Borated Aluminum Material Prolonged exposure to elevated temperatures can lead to a loss of fracture toughness and ductility in some materials as a result of changes to their microstructure. Testing of aluminum-based neutron poison plates, however, has shown that these materials typically increase in ductility when they are aged at high temperatures. Qualification tests performed on neutron poisons demonstrate that microstructural changes induced by aging typically make the aluminum softer and more ductile as it is annealed, while the boron and carbide particulates are thermally stable at TN-68 cask internal temperatures.

Also, as discussed in Section 3.4.3.6.4 above, the borated aluminum subcomponents do not serve a structural function and are supported by adjacent subcomponents.

Therefore, decreases in strength due to thermal aging are not expected to affect the intended safety functions of the borated aluminum subcomponents.

Therefore, thermal aging of the borated aluminum material is not credible in a helium environment.

3.4.3.6.6 Radiation Embrittlement of Borated Aluminum Material Embrittlement of metals may occur under exposure to neutron radiation. Depending on the neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, reduced fracture toughness, and decreased resistance to cracking.

Section 3.4.2.7 of NUREG-2214 [3-2] describes data showing that the mechanical properties of aluminum are not degraded at neutron fluence levels on the order of 1020 neutrons/square centimeter (n/cm2). Section 3.2.1.9 of NUREG-2214 describes a bounding calculation that showed estimated fluence level within a dry storage cask system is four orders of magnitude below the levels reported to degrade the mechanical properties of aluminum-based neutron poison.

Therefore, radiation embrittlement of the borated aluminum material is not credible in a helium environment.

3.4.3.6.7 Wet Corrosion and Blistering of Borated Aluminum Material The core of aluminum-boron carbide laminate composites is not fully sintered and, as a result, can have a porosity of 1 to 8 percent with varying degrees of interconnectivity among pores. This may allow water ingress into the core, where the water can react with the aluminum to form aluminum oxide and hydrogen gas. Blistering has been experimentally observed in the Boral cladding during repeated wetting and drying representative of dry storage operations. Tests simulating the wetting and vacuum drying cycles during canister closure operations show that Boral can form blisters in the aluminum cladding because of water ingress through its exposed edges. The blisters are characterized by a local area where the aluminum cladding separates from the underlying boron carbide-aluminum core, and the cladding is physically deformed outward.

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CoC 1027 Renewal Application Revision 0 / April 2020 There is no OE recorded in the United States, however, on the blistering of aluminum-boron carbide laminate composites used in loaded dry storage systems (DSSs).

Although indications of wet corrosion have been observed during DSS loadings (via indirect measurements of generated hydrogen), this aging mechanism has not been observed to result in blistering or reduce the neutron-absorbing capability of Boral in loaded DSSs. Further, because only a trace amount of water will be left in a dry storage cask after dehydration and helium backfill, wet corrosion is not expected to compromise the intended function of aluminum-boron carbide laminate composites during the PEO.

Therefore, wet corrosion and blistering of the borated aluminum material is not credible in a helium environment.

3.4.3.7 Aging Mechanism of Concrete Material A review of Table 3-5 through Table 3-8 shows that the environments that the concrete subcomponents are exposed to are:

  • Air-outdoor
  • Groundwater/soil The following aging mechanisms for concrete material were evaluated to determine if they are credible in the environments that concrete is exposed to:
  • Freeze-thaw
  • Creep
  • Reaction with aggregates
  • Differential settlement
  • Aggressive chemical attack
  • Corrosion of reinforcing steel
  • Shrinkage
  • Radiation damage
  • Fatigue
  • Dehydration at high temperature
  • Microbiological degradation
  • Delayed ettringite formation
  • Salt scaling
  • Hardening Page 3-35

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.7.1 Freeze-Thaw of Concrete Material Concretes that are nearly or fully saturated with water can be damaged by repeated freezing and thawing cycles. The degradation mode would initiate at the outer concrete surface exposed to outdoor environments, primarily at horizontal surfaces where water ponding can occur. For below-grade concrete structures, water that resides in soil can also be subject to freezing conditions, potentially promoting freeze-thaw damage. Because water expands when freezing, fully or mostly saturated concrete will experience internal stresses from the expanding ice, which can cause concrete cracking or scaling when pressures exceed the concrete tensile strength.

Therefore, freeze-thaw of concrete material is considered credible in air-outdoor or groundwater/soil environments.

3.4.3.7.2 Creep of Concrete Material Creep in concrete is the time-dependent deformation resulting from sustained loads.

While there are several factors that affect creep in concrete, the most important parameter controlling creep is sustained loading. The creep rate in concrete decreases exponentially with time, i.e., creep would be more of a concern during the initial license period rather than the PEO. In addition, the creep in concrete is largely mitigated by proper design practices. Furthermore, creep-induced concrete cracks are not generally large enough to reduce the compressive strength of concrete, cause deterioration of concrete, or cause exposure of reinforcing steel to the environment.

Therefore, creep of concrete material is not credible in air-outdoor or groundwater/soil environments.

3.4.3.7.3 Reaction with Aggregates of Concrete Material The two most common alkali-aggregate reactions are alkali-silica reaction (ASR) and alkali-carbonate reaction, with ASR being the most common and damaging. ASR is a chemical reaction between hydroxyl ions (present in the alkaline cement pore solution) and reactive forms of silica present in some aggregates. ASR damage in the concrete manifests itself as a characteristic map cracking on the concrete surface. The internal damage results in the degradation of concrete mechanical properties and, in severe cases, the expansion can result in undesirable dimensional changes and popouts. In general, ASR is a slow degradation mechanism that can cause serviceability issues and may exacerbate other deterioration mechanisms. The requisite conditions for initiation and propagation of ASR include 1) a sufficiently high alkali content of the cement (or alkali from other sources, such as deicing salts, seawater, and groundwater), 2) a reactive aggregate, and 3) available moisture, generally accepted to be relative humidity greater than 80 percent.

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CoC 1027 Renewal Application Revision 0 / April 2020 ASR may take from three to more than 25 years to develop in concrete structures, depending on the nature (reactivity level) of the aggregates, the moisture and temperature conditions to which the structures are exposed, and the concrete alkali content. The delay in exhibiting deterioration indicates that there may be less reactive forms of silica that can eventually cause deterioration. Recent OE has revealed degradation of the concrete at a nuclear power plant as a result of ASR even though the concrete used at the plant passed all industry standard ASR screening tests at the time of construction. In addition, ASR screening tests are not conducted on each aggregate source but rather in select batches.

Because of the uncertainties in screening tests that can effectively be used to eliminate the potential for ASR and previous ASR OE at a nuclear facility, the aging mechanism is considered credible in concrete exposed to any environment with available moisture.

Therefore, reaction with aggregates of concrete material is considered credible in air-outdoor or groundwater/soil environments.

3.4.3.7.4 Differential Settlement of Concrete Material Differential settlement is a result of the uneven deformation of the supporting foundation soil. Differential settlement, which causes distortion (loss of form) and damage (cracking) to concrete structures, is a function of the uniformity of the soil, stiffness of the structure, stiffness of the soil, and distribution of loads within the structure. The settlement of saturated cohesive soil consists of three components:

1) immediate settlement occurring due to the applied load, 2) consolidation settlement occurring gradually due to dissipation of the excess pore pressures generated by the applied load, and 3) secondary compression that depends on the composition and structure of the soil skeleton. The settlement of course-grained granular soils subject to applied load occurs immediately, primarily from the compression of the soil skeleton due to rearrangement of particles. However, most settlement issues involving a combination of immediate settlement and progressing long-term settlement are typically discovered in less than one year of construction. However, OE has shown that it can occur. Therefore, differential settlement of concrete exposed to groundwater or soil (below-grade) environment is considered credible.

Although portions of concrete structures in contact with air can exhibit the effects of settlement, the direct interaction with the underlying soil is the mechanism that causes the aging. As a result, differential settlement of concrete exposed to an outdoor air environment is not considered credible.

Therefore, differential settlement of concrete material is considered credible in a groundwater/soil environment, but not credible in an air-outdoor environment.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.7.5 Aggressive Chemical Attack of Concrete Material The intrusion of aggressive ions or acids into the pore network of the concrete can cause various degradation phenomena. The aggressive chemical attack typically originates from an external source of sulfate or magnesium ions as well as acidic environmental conditions.

In an air-outdoor or groundwater/soil environment, groundwater, seawater, and rainwater may contain sulfate species that penetrate the concrete and chemically react with alkali and calcium ions to form a precipitate of calcium sulfate in addition to other forms of calcium and sulfate-based compounds. The manifestation of sulfate attack is cracking, increase in concrete porosity and permeability, loss of strength, and surface scaling generated by the expansion associated with the formation of ettringite within the concrete and the pressure generated by the precipitated calcium and sulfate-base compounds inside the concrete pore network. Acids from groundwater and acid rain can dissolve both hydrated and unhydrated cement compounds (e.g., calcium hydroxide, calcium silicate hydrates, and calcium aluminate hydrates) as well as calcareous aggregate in concrete without any significant expansion reaction. In most cases, the chemical reaction forms water-soluble calcium compounds, which are then leached away by aqueous solutions. The dissolution of concrete commences at the surface and propagates inward as the concrete degrades. The signs of acidic attack are loss of alkalinity (also disturbing of electrochemical passive conditions for the embedded steel reinforcement), loss of material (i.e., concrete cover), and loss of strength.

NUREG-1801, [3-4], states that continued or frequent cyclic exposure to the following aggressive chemical environments is necessary to cause significant aggressive chemical attack degradation:

  • Acidic solutions with pH < 5.5
  • Sulfate solutions > 1500 ppm Since the groundwater/soil and air-outdoor may contain solutions that exceed these criteria, aggressive chemical attack is considered possible.

Therefore, aggressive chemical attack of concrete material is considered credible in air-outdoor or groundwater/soil environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.7.6 Corrosion of Reinforcing Steel of Concrete Material Corrosion of the reinforcing steel embedded in the concrete is mainly caused by the presence of chloride ions in the concrete pore solution and carbonation of the concrete.

The highly alkaline environment provided by the concrete (normally with porewater pH > 13.0) results in the formation of a metal-adherent oxide film on the reinforcement steel bar surface, which passivates the steel. However, chloride ions may penetrate the concrete matrix and break down the steel passive layer, once the chloride concentration at the reinforcing steel surface exceeds a threshold value, triggering corrosion of the reinforcing steel and shortening the service life of a concrete structure. Concrete durability is directly related to the quality of the concrete, the external concentration of chlorides on the concrete surface, and the reinforcement material. The service life of concretes exposed to chloride attack depends on the concrete cover, the surface chloride concentration, the chloride diffusion coefficient, the type of cementitious material, and the reinforcing steel material.

In an air-outdoor or groundwater/soil environment, chloride ions may penetrate the concrete from the outside environment, such as when using deicing salts, aggressive groundwater, or marine environments. Although no cases of corrosion-induced damage have been reported, the results of a durability model show that corrosion of the reinforcing steel in concrete can potentially initiate and propagate within the 60-year timeframe for concretes of moderate to low quality.

Therefore, corrosion of reinforcing steel of concrete material is considered credible in air-outdoor or groundwater/soil environments.

3.4.3.7.7 Shrinkage of Concrete Material Shrinkage occurs when hardened concrete dries from a saturated condition to a state of equilibrium in about 50 percent relative humidity. As excess concrete water evaporates, tensile stresses are induced in the concrete due to internal pressure from the capillary action of water movement, which results in cracking. The factors affecting shrinkage are cement content, water-to-cement ratio, degree of hydration, elastic modulus of aggregates, amount and characteristics of concrete admixtures, temperature and humidity during curing, and size and shape of concrete.

Shrinkage of concrete occurs initially during curing, which can be controlled through concrete formulation and the density and distribution of internal reinforcement. Over 90 percent of the shrinkage occurs during the first year, reaching 98 percent by the end of the first five years. Thus, shrinkage is not expected to influence concrete performance during the PEO, because most of the shrinkage will take place early on in the life of the concrete. As a result, shrinkage of concretes is not considered credible.

Therefore, shrinkage of concrete material is not credible in air-outdoor or groundwater/soil environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.7.8 Leaching of Calcium Hydroxide of Concrete Material A constant or intermittent flux of water over a concrete surface can result in the removal or leaching of calcium hydroxide. Calcium hydroxide leaching is observed in the form of white leachate deposits (calcium carbonate) on the concrete surface. The extent of the leaching depends on the environmental salt content and temperature, and it can take place above and below ground.

In an air-outdoor or groundwater/soil environment, the concrete surface may be exposed to intermittent fluxes of water, thus making leaching of calcium hydroxide credible.

Therefore, leaching of calcium hydroxide of concrete material is considered credible in air-outdoor or groundwater/soil environments.

3.4.3.7.9 Radiation Damage of Concrete Material Radiation effects on concrete properties will depend on the gamma and neutron radiation doses, temperature, and exposure period. Gamma radiation can decompose and evaporate water in concrete. Because most of the water is contained in the cement paste, the effect of gamma radiation on cement paste is more significant than on the aggregates. Gamma radiation can also decompose the Si-O bond within calcium silicate hydrate. Neutron radiation deteriorates concrete by reducing stiffness, forming cracks by swelling, and changing the microstructure of the aggregates.

Section 3.5.1.9 of NUREG-2214 [3-2] describes data showing that the mechanical properties of concrete are not degraded at neutron fluence levels on the order of 1019 neutrons/square centimeter (n/cm2) and 1010 rad for gamma rays. Section 3.2.1.9 of NUREG-2214 describes a bounding calculation that showed estimated neutron fluence level within a dry storage cask system is three orders of magnitude below the levels reported to degrade the mechanical properties of concrete. The gamma dose is also expected to be several orders of magnitude less than the 1010 rad limit.

Therefore, radiation damage of concrete material is not credible in air-outdoor or groundwater/soil environment.

3.4.3.7.10 Fatigue of Concrete Material Concrete fatigue strength is defined as the maximum stress that the concrete can sustain without failure under a given number of stress cycles. Because the storage pad is a static application, mechanical cyclic loading is not expected. However, restraint of the concrete from expanding and contracting as it is exposed to rapid changes in temperature will lead to internal stresses in the structure. If the changes in temperature are severe and the resulting strains are sufficient, local plastic deformation can occur.

Repeated application of this thermal loading can lead to crack initiation and propagation in low-cycle fatigue. Concrete fatigue in the reinforced concrete storage pad may be caused by diurnal and seasonal temperature gradients.

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CoC 1027 Renewal Application Revision 0 / April 2020 Section 3.5.1.10 of NUREG-2214 [3-2] discusses a generic evaluation of concrete fatigue over 60 years of storage considering an extreme seasonal ambient temperature variation from -40 °F to 125 °F. This range bounds the design temperature range for the TN-68 dry storage cask system of -20 °F to 100 °F. The evaluation concluded that fatigue of concrete exposed to sheltered, outdoor, groundwater or soil (below-grade),

and fully encased environments is not considered credible.

Therefore, fatigue of concrete material is not credible in air-outdoor or groundwater/soil environments.

3.4.3.7.11 Dehydration at High Temperature of Concrete Material Exposure of concrete to elevated temperatures can affect its mechanical and physical properties. It is well known that concretes can degrade at high temperatures due to dehydration of the hydrated cement paste, thermal incompatibility between the cement and aggregates, and physicochemical deterioration of the aggregates.

The effects of thermal dehydration were addressed during the initial licensing of the TN-68 dry storage cask system. Since the fuel temperature decreases over time, the concrete temperature will also decrease over time. Therefore, the design temperature considerations during the initial licensing are expected to remain adequate during the PEO.

Therefore, dehydration at high temperature of concrete material is not credible in air-outdoor or groundwater/soil environments.

3.4.3.7.12 Microbiological Degradation of Concrete Material Biodeterioration is caused by colonization of microbes and microorganisms that grow on concrete surfaces that offer favorable environmental conditions (e.g., available moisture, near neutral pH, presence of nutrients). Conducive environments may have elevated relative humidity (i.e., greater than about 60 percent), long cycles of humidification and drying, freezing and thawing, high carbon dioxide concentrations, high concentrations of chloride ions or other salts, or high concentrations of sulfates and small amounts of acids. Biodeterioration may lead to reduction of the protective cover depth and increase both concrete porosity and the transport of aggressive chemicals. In addition, this degradation mode can promote a reduction in concrete pH, loss of concrete strength, and spalling/scaling. Evidence shows that a wide variety of organisms can cause concrete deterioration in polluted soils and groundwater. The biodeterioration of concrete typically is confined to the surface. The rate of deterioration is slow, but the degradation mode has been observed within 40 years of exposure.

Although no cases of microbiological degradation of concrete have been reported in nuclear applications, the degradation mode is considered credible, as below-grade environments may be conducive to microbe and bacteria growth.

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CoC 1027 Renewal Application Revision 0 / April 2020 In air-outdoor environments, favorable conditions for microbiological degradation mechanisms may exist because of the potential presence of moisture. However, the conditions would be intermittent, and there is no evidence that actual concrete subcomponents in these environments microbiologically degrade.

Therefore, microbiological degradation of concrete material is considered credible in a groundwater/soil environment, but is not credible in an air-outdoor environment.

3.4.3.7.13 Delayed Ettringite Formation of Concrete Material At the initial stage of fresh concrete curing, ettringite, commonly referred to as naturally occurring ettringite, is formed by the reaction of tricalcium aluminate and gypsum in the presence of water. The formation of naturally occurring ettringite in fresh concrete is not detrimental to the overall concrete performance. At the still-early stage of concrete curing, the naturally occurring ettringite may convert to monosulfoaluminate if curing temperatures are greater than about 70 °C (158 °F).

After concrete hardens, if the temperature decreases below this value, the monosulfoaluminate becomes unstable and, in the presence of sulfates released by the calcium-silicate-hydrate gel, ettringite will reform. This mechanism is called delayed ettringite formation (DEF), which results in volume expansion and increased internal pressures in the concrete.

The conditions necessary for the occurrence of DEF are excessive temperatures during concrete placement and curing, the presence of internal sulfates, and a moist environment. American Concrete Institute (ACI) 318-05 indicates that inspection reports shall document concrete temperature and protection during placement when the ambient temperature is above 35 °C (95 °F). Protection measures during concrete placement include lowering the temperature of the batch water, cement, and aggregates as referenced in ACI 305R-10. As such, following the ACI 318-05, ACI 305R-10, and ACI 308R-01 guidelines during concrete placement and curing can effectively limit the concrete temperature to below 70 °C (158 °F), therefore preventing the development of DEF.

As noted in Section 3.3.2, the storage pad is designed and constructed in accordance with codes and standards set by the general licensee. Construction practices for the concrete storage pad may vary from ISFSI to ISFSI. Therefore, DEF may be an applicable concrete aging mechanism for the storage pad, unless ruled out by the general licensee based on an ISFSI-specific evaluation.

Therefore, delayed ettringite formation of the storage pad concrete material is considered credible in air-outdoor or groundwater/soil environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.7.14 Salt Scaling of Concrete Material Salt scaling is defined as superficial damage caused by freezing a saline solution on the surface of a concrete body. The damage is progressive and consists of the removal of small chips or flakes of material. Similar to freeze-thaw damage, salt scaling takes place when concrete is exposed to freezing temperatures, moisture, and dissolved salts. The degradation is maximized at a moderate concentration of salt (e.g., from deicing salts).

Similar to freeze-thaw damage, the degradation initiates at the outer concrete surface exposed to outdoor environments, primarily at horizontal surfaces where water ponding can occur. For below-grade concrete structures, water that resides in soil may contain salt and, thus, the concrete may be susceptible to salt scaling.

Therefore, salt scaling of concrete material is considered credible in air-outdoor or groundwater/soil environments.

3.4.3.7.15 Hardening of Concrete Material Concrete compressive strength is known to increase with age. As the compressive strength increases, the concrete hardens also increases causing the applicable deceleration g-loads during cask drops and tip-over accidents to increase. This could potentially affect the structural capability of the TN-68 casks.

UFSAR [3-3] Section 3D.3.2 describes the inputs used for the concrete material (i.e.,

material law 16 in LS-DYNA) in the cask tip over analysis, per the design basis calculation, these inputs represent 4,200 psi compressive strength concrete. UFSAR Section 3D.7.2 lists the concrete strength used in the cask end drop analysis as 6,000 psi.

The data in Table 2.13 of NUREG/CR-6424 [3-8], shows the time-dependent compressive strength values for various Portland cement concretes at 28 days, 20 years, and 30 years. The table shows that concrete with a higher initial compressive strength will realize a smaller percentage increase in strength over time, versus concrete with a lower initial compressive strength. For example, concrete with a 28-day compressive strength 19.2 MPa will increase 87% during the first 20 years and additional 10% during the next 10 years. While concrete with a 28-day compressive strength of 33.6MPa will increase only 41% during the first 20 years and a negligible amount beyond 20 years. Since the minimum design basis concrete compressive strength for the CoC 1027 concrete storage pads is 4,200 psi (28.9 Mpa), it is reasonable to conclude that any increase in the concrete hardness will occur during the first 20 years of storage and not during the PEO.

Therefore, hardening of concrete material is not credible in air-outdoor or groundwater/soil environments.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.8 Aging Mechanism of Spent Fuel Assembly Cladding A review of Table 3-5 through Table 3-8 shows that the environment that the spent fuel assembly (SFA) cladding is exposed to is:

  • Helium The following aging mechanisms for SFA cladding were evaluated to determine if they are credible in the environment that SFA cladding is exposed to:
  • Hydride-induced embrittlement
  • Delayed hydride cracking
  • Thermal creep
  • Low temperature creep
  • Mechanical overload
  • Oxidation
  • Pitting corrosion
  • Galvanic corrosion
  • Radiation embrittlement
  • Fatigue 3.4.3.8.1 Hydride-Induced Embrittlement of Spent Fuel Assembly Cladding In reactor service, the zirconium-based fuel cladding absorbs hydrogen, which leads to the precipitation of hydride platelets as the dissolved hydrogen exceeds the solubility limit of the cladding. The primary source of the hydrogen is water-side corrosion (oxidation) of the cladding. The total concentration of hydrogen absorbed by the cladding (i.e., dissolved in the zirconium matrix and in precipitated hydrides) increases with burnup and varies axially across the fuel rods. When discharged from the reactor and during wet storage, the hydride platelets are mostly oriented in the circumferential-axial direction, with a smaller fraction oriented in the radial-axial direction.

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CoC 1027 Renewal Application Revision 0 / April 2020 During vacuum drying of the TN-68 casks, the temperature of the SFAs and the temperature-dependent solubility limit of hydrogen in the cladding will also increase.

As a result, some of the hydrides present in the cladding will redissolve as hydrogen.

Once the loaded cask is dried and backfilled, the cladding temperature will decrease over time and upon a sufficient temperature drop, some of the hydrogen in solution will reprecipitate as new hydrides. During this process, the orientation of these precipitated hydrides may change from the circumferential-axial to the radial-axial direction. The degree of reorientation is primarily driven by the metallurgical microstructure of the cladding alloy and the cladding hoop stresses during drying operations and subsequent cooling, which are determined by the rod internal pressure at a given gas temperature. Cladding with a high concentration of radial hydrides (determined by the dry storage cask system drying conditions) has been shown to have reduced ductility under pinch-load stresses at sufficiently low temperatures. Section 3.6.1.1 of NUREG-2214 [3-2] contains more discussion of the hydride reorientation phenomenon, including expected ranges of dissolved hydrogen, drying temperatures, solubility limits, hoop stresses, and experimental studies.

Considering the hydrogen content, peak drying temperatures, and corresponding hoop stresses, hydride reorientation in zirconium-based cladding is only credible for high burnup (HBU) fuel during the 60-year service timeframe. Furthermore, depending on the specific fuel contents, it is possible for some of the cladding to reach temperatures near or below the ductile-to-brittle transition temperatures reported in the literature.

Therefore, degradation of mechanical properties during pinch-type stresses due to hydride-induced embrittlement is considered a credible aging mechanism for zirconium-based HBU (i.e., > 45 GWd/MTU) fuel claddings. Hydride-induced embrittlement is not credible for low burnup (i.e., 45 GWd/MTU) fuel.

3.4.3.8.2 Delayed Hydride Cracking of Spent Fuel Assembly Cladding Delayed hydride cracking (DHC) is a time-dependent mechanism traditionally thought to occur by the diffusion of hydrogen to an incipient crack tip (notch, flaw) in the cladding, followed by nucleation, growth, and subsequent fracture of the precipitated hydrides at the crack tip. Hydrogen dissolved in the cladding (see Section 3.4.3.8.1) can diffuse up a stress gradient in the crystalline lattice, or into the stress field at the core of an edge dislocation. The concentration gradient established by the stress gradient may lead to hydrogen supersaturation (i.e., solubility limit being exceeded) leading to the precipitation of hydrides at the crack tip. The precipitated hydride will continue to grow by the dissolution of hydrides in the low-stress regions of the material and by the continued diffusion of hydrogen up the stress gradient. Once the hydride reaches a critical size, it will crack and propagate to the end of the hydride, where it will blunt. The cycle could then repeat, until the crack propagates through the thickness of the material.

Page 3-45

CoC 1027 Renewal Application Revision 0 / April 2020 Requisite conditions for DHC are the presence of: 1) hydrides, 2) existing crack tips (notch, flaws) that act as initiating sites, and 3) sufficient cladding hoop stresses.

Section 3.6.1.2 of NUREG-2214 [3-2] discusses the availability of hydrides, initial depth of existing crack tips, and hoop stresses in zirconium-based fuel claddings.

NUREG-2214 goes on to determine that the critical flaw size needed to initiate DHC is larger than the initial depth of potentially existing cracks.

Therefore, DHC is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.3 Thermal Creep of Spent Fuel Assembly Cladding Creep is the time-dependent deformation of a material under stress. Creep in zirconium-based cladding is caused by the hoop stresses from the rod internal pressure at a given fuel temperature. Therefore, the mechanism is expected to be self-limiting, due to the decreasing temperatures and creep-induced volume expansion, which results in lower internal rod pressures with time. Excessive creep of the cladding during dry storage could lead to thinning, hairline cracks, or gross ruptures. Section 3.6.1.3 of NUREG-2214 [3-2] contains an evaluation of cladding hoop stresses during dry storage for HBU fuel. The evaluation estimated the maximum cladding strain to be near 2.1 percent. The elastic strain limit for various zirconium-based cladding alloys with circumferential hydrides is less than 1 percent and is expected to be lower for cladding containing both circumferential and radial hydrides. Therefore, zirconium-based cladding in HBU fuel is expected to undergo creep strains during the 60-year timeframe.

Interim Staff Guidance (ISG) -11, Revision 3 [3-6], provides the basis for which thermal creep of low burnup (i.e., 45GWd/MTU) zirconium-based cladding is not a concern provided the peak normal cladding temperature is below 752 °F. Table 3-3 shows the maximum cladding temperature for normal condition for fuel stored in a TN-68 cask is 649°F. Therefore, creep of low burnup fuel stored in a TN-68 cask is not a concern.

Therefore, thermal creep is considered a credible aging mechanism for zirconium-based HBU (i.e., > 45 GWd/MTU) fuel claddings. Thermal creep is not credible for low burnup (i.e., 45 GWd/MTU) fuel with stainless steel or zirconium-based alloy cladding.

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CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.8.4 Low Temperature Creep of Spent Fuel Assembly Cladding Low-temperature creep (also called athermal creep) may occur when sustained hoop stresses operate on the cladding material at or near ambient temperature. Various athermal creep mechanisms have been proposed at low stresses, although there is no evidence or literature information to support that these will be operational on zirconium-based alloys. However, the literature shows that low-temperature creep has been shown to occur in titanium and its alloys, which leads to deformation twinning.

Since both titanium and zirconium have the same crystalline structure (hexagonal close packed crystalline), Section 3.6.1.4 of NUREG-2214 [3-2] performed an evaluation using the titanium threshold stress for low-temperature creep of 25 percent of the yield strength as the threshold stress for zirconium-based cladding.

The evaluation concluded that the room temperature hoop stresses on the zirconium-based cladding are expected to be less than 25 percent of the yield strength during the 60-year storage timeframe.

Therefore, low temperature creep is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.5 Mechanical Overload of Spent Fuel Assembly Cladding Mechanical overload is generally associated with pellet-to-cladding mechanical interaction (PCMI), which could compromise the cladding integrity during storage.

PCMI is likely during reactor operations when the reactivity transient during a reactivity-initiated accident results in a rapid increase in a fuel rod power, leading to a nearly adiabatic heating of the fuel pellets and potential failure of the fuel cladding.

Data generated in experimental reactors conducting ramp testing of heavily hydrided fuel claddings (i.e., associated with HBU fuel) indicate that hydride rims with large hydride number density at the cladding outer surface may lead to crack initiation. The cracks could propagate from the outside toward the inner cladding surface, potentially resulting in failures. During dry storage, PCMI stresses could develop due to pellet swelling and release of fission gases to the gap between the fuel and cladding.

Section 3.6.1.5 of NUREG-2214 [3-2] contains a generic evaluation showing that the strain rates during storage of HBU zirconium-based cladding are expected to be five to seven orders of magnitude lower than the threshold strain rate needed for PCMI-induced failures.

Due to the expected lower hydride concentration in low burnup fuel ( 45 GWd/MTU), PCMI-induced failures are not considered credible.

Therefore, mechanical overload is not credible for zirconium-based alloy SFA cladding.

Page 3-47

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.8.6 Oxidation of Spent Fuel Assembly Cladding There is very little residual water in the internal environments of a TN-68 cask following drying and refilling with inert helium gas, the oxidation of the external surface of the cladding will be limited. Similarly, there is no moisture internal to the fuel pin that could contribute to the oxidation of the cladding.

Therefore, oxidation is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.7 Pitting Corrosion of Spent Fuel Assembly Cladding Because there is very little residual water in the internal environments of a TN-68 cask following drying and refilling with inert helium gas, pitting corrosion of the external surface of the cladding will be limited. Similarly, there is no moisture internal to the fuel pin that could contribute to pitting corrosion of the cladding.

Therefore, pitting corrosion is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.8 Galvanic Corrosion of Spent Fuel Assembly Cladding Galvanic corrosion can occur due to a mismatch in corrosion potentials between two metals in an aqueous solution. In fuel assemblies, the mismatch can occur when the cladding is in contact with other metallic components, which could result in the formation of a galvanic cell, provided there is an aqueous solution between the two subcomponents.

Because there is very little residual water in the internal environments of a TN-68 cask following drying and refilling with inert helium gas, the oxidation of the external surface of the cladding will be limited. Similarly, there is no moisture internal to the fuel pin that could contribute to the corrosion of the cladding.

Therefore, galvanic corrosion is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.9 Stress Corrosion Cracking of Spent Fuel Assembly Cladding Stress corrosion cracking occurs as a result of a synergistic combination of a susceptible material, an aggressive environment, and sufficiently high tensile stress.

The corrosive environment associated with SCC of fuel rods has been attributed to specific fission products, such as iodine, cesium, and cadmium, generated during reactor irradiation. Stress corrosion cracking of the cladding can occur at the inner surface of the rod where the fuel pellet and cladding mechanically interact, and is related to PCMI hoop stresses on the cladding.

Section 3.6.1.9 of NUREG-2214 [3-2] contains an evaluation showing that hoop stresses during a 60-year period of storage, including those due to PCMI, will remain below that needed for inducing SCC in zirconium-based cladding.

Therefore, SCC is not credible for zirconium-based alloy SFA cladding.

Page 3-48

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.8.10 Radiation Embrittlement of Spent Fuel Assembly Cladding Radiation embrittlement of cladding can result in degradation of the mechanical properties of the cladding, such as ductility and strength. Embrittlement is largely observed during reactor operation due to cumulative fast neutron fluence on the order of 1022 n/cm2.

Section 3.2.1.9 of NUREG-2214 [3-2] describes a bounding calculation that showed estimated fluence after 100 years of storage, which is five to six orders of magnitude below the fluence seen during reactor operations. Therefore, 100 years of storage has a negligible contribution to the overall fast neutron fluence.

Therefore, radiation embrittlement is not credible for zirconium-based alloy SFA cladding.

3.4.3.8.11 Fatigue of Spent Fuel Assembly Cladding Fatigue is the progressive and localized structural damage that occurs when a material is subjected to cyclic loading. The only cyclic loading experienced by the spent fuel cladding is associated with thermal cycling. The only source of potential thermal fatigue of the spent fuel cladding is ambient seasonal and daily temperature fluctuation. The spent fuel cladding does not experience the full amplitude of ambient temperature cycles, and a gradual, long-term temperature decrease occurs during the course of storage. The seasonal and daily variations in ambient conditions are ameliorated by the thermal mass of the TN-68 cask.

Section 3.6.1.11 of NUREG-2214 [3-2] contains a very conservative evaluation of the cyclic stresses in zirconium-based spent fuel cladding due to thermal cycles over a 60-year storage period. It concludes that the cumulative cyclic stresses are well below the threshold needed for fatigue-induced failure of the zirconium-based cladding.

Therefore, fatigue is not credible for zirconium-based alloy SFA cladding.

3.4.3.9 Aging Mechanism of Spent Fuel Assembly Hardware Materials A review of Table 3-5 through Table 3-8 shows that the environment that the SFA hardware materials are exposed to is:

  • Helium The following aging mechanisms for SFA hardware materials were evaluated to determine if they are credible in the environment that SFA hardware materials are exposed to:
  • Creep
  • Hydriding
  • General corrosion Page 3-49

CoC 1027 Renewal Application Revision 0 / April 2020

  • Radiation embrittlement
  • Fatigue 3.4.3.9.1 Creep of Spent Fuel Assembly Hardware Materials As discussed in Section 3.4.3.1.6, as a general rule of thumb, at temperatures below 0.4Tm, where Tm is the melting point of the metal in Kelvin, thermal activation is insufficient to produce creep that could compromise the intended functions of SSCs.

The melting temperature of various zirconium alloys, Inconel, and stainless steel alloys is above 3,272 °F, 2,300 °F, and 2,597 °F, respectively. Appling the 0.4Tm criterion yields a creep threshold of 1,033 °F for zirconium alloys, 644 °F for Inconel alloys, and 763 °F for stainless steel.

The highest temperatures within the TN-68 dry storage cask system are at locations close to the fuel rods. As shown in Table 3-3 the maximum fuel cladding temperature was determined to be 649 °F. Because the fuel rods are the only heat source within the system, they provide upper temperature limits for all subcomponents regardless of their environment. It is apparent from the temperatures in Table 3-3 that SFA hardware made of zirconium alloys and stainless steel will not approach the minimum 1,033 °F and 763 °F creep threshold temperatures, respectively. While the maximum cladding temperature of 649 °F is above the 644 °F Inconel alloy threshold temperature, it will decrease during the initial storage period such that it will be below the Inconel threshold temperature during the PEO.

Therefore, creep is not credible for the SFA hardware materials.

3.4.3.9.2 Hydriding of Spent Fuel Assembly Hardware Materials Hydriding may occur in zirconium alloys that experience hydrogen pickup in reactor service. As the temperature of the assembly hardware decreases, zirconium hydrides precipitate due to the decreasing hydrogen solubility in the zirconium matrix. The hydride precipitation will occur when the hardware cools in the spent fuel pools after reactor discharge. Some of the hydride will dissolve during the drying process and will reprecipitate due to subsequent cooling during storage. Unlike the fuel rod cladding, there is no hoop stress for the zirconium-based assembly hardware to cause hydride reorientation. Any load on the assembly hardware is predominantly due to its own weight, which is not sufficient to cause hydride reorientation. Because there is limited load during storage on assembly hardware, it is unlikely that hydriding will affect the ability of the assembly hardware to ensure that the spent fuel remains in the as-analyzed configuration.

Hydride formation is not a concern in Inconel and stainless steel materials, thus hydriding is not credible in stainless steel spent fuel assembly hardware.

Therefore, hydriding is not credible for the SFA hardware materials.

Page 3-50

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.9.3 General Corrosion of Spent Fuel Assembly Hardware Materials There is very little residual water in the internal environments of a TN-68 cask following drying and refilling with inert helium gas, thus the general corrosion of the SFA hardware will be limited.

Therefore, general corrosion is not credible for the SFA hardware materials.

3.4.3.9.4 Stress Corrosion Cracking of Spent Fuel Assembly Hardware Materials SCC of SFA hardware made of zirconium alloys is not considered credible based on the discussion in Section 3.4.3.8.9.

Various stainless steel and Inconel assembly hardware components could be susceptible to SCC in the presence of an aggressive environment and sufficient residual tensile stresses. Residual tensile stresses are expected to be present in the assembly hardware, primarily in welded areas. Regarding the chemical environment, the SFA hardware is exposed to the inert helium environment within the TN-68 cask.

In addition there is very little residual water in the internal environments of a TN-68 cask following drying and refilling with inert helium gas. Because of the lack of halides and the small amount of water in the helium environments, SCC of stainless steel and Inconel hardware components is not considered credible.

Therefore, SCC is not credible for the SFA hardware materials.

3.4.3.9.5 Radiation Embrittlement of Spent Fuel Assembly Hardware Materials Radiation embrittlement of SFA hardware made from zirconium alloys is excluded using the basis provided in Section 3.4.3.8.10. This basis (i.e., the fluence seen during dry storage is several orders of magnitude below that seen during reactor operation) is also applicable to SFA hardware made of Inconel. Similarly, radiation embrittlement of assembly hardware made of stainless steel is not considered credible per the technical bases provided in Section 3.4.3.2.9.

Therefore, radiation embrittlement is not credible for the SFA hardware materials.

3.4.3.9.6 Fatigue of Spent Fuel Assembly Hardware Materials Fatigue of SFA hardware made from zirconium alloys is excluded using the basis provided in Section 3.4.3.8.11. Fatigue of assembly hardware made of stainless steel is not considered credible per the technical basis provided in Section 3.4.3.2.7. Similarly rational (as used for zirconium and stainless steel materials) may also be used to show that fatigue of assembly hardware made of Inconel is not considered credible.

Therefore, fatigue is not credible for the SFA hardware materials.

Page 3-51

CoC 1027 Renewal Application Revision 0 / April 2020 3.4.3.10 Summary of Aging Mechanism of Materials Table 3-4 provides a summary of the aging mechanisms for materials used in the TN-68 dry storage cask system. The table also lists the environments where the aging mechanism could manifest into an aging effect that affects an important-to-safety function, i.e., is considered credible. Table 3-4 also identifies the aging effect into which each credible aging mechanism could manifest, (based on Table 2-4 of NUREG-2214 [3-2]).

Page 3-52

CoC 1027 Renewal Application Revision 0 / April 2020 3.5 Aging Management Review for TN-68 Cask Table 3-5 provides the detailed results of the AMR for the subcomponent parts of the TN-68 cask. For each material group/environment combination and each part, the table lists the credible aging mechanisms and effects based on the information in Table 3-4. A summary of the aging effects that require management for the TN-68 cask is provided in Table 3-6.

The following aging effects/mechanisms will be managed via the TN-68 AMP:

  • Steel Loss of material due to general, pitting, crevice, and galvanic corrosion Loss of preload due to stress relaxation of bolts
  • Stainless steel Loss of material due to pitting, crevice, and galvanic corrosion
  • Polymers Shrinkage/cracking due to thermal aging and radiation embrittlement
  • Aluminum Loss of material due to general, pitting, crevice, and galvanic corrosion Page 3-53

CoC 1027 Renewal Application Revision 0 / April 2020 3.6 Aging Management Review for Storage Pad Table 3-7 provides the detailed results of the AMR for the subcomponent parts of the storage pad. For each material group/environment combination and each part, the table lists the credible aging mechanisms and effects based on the information in Table 3-4.

The following aging effects/mechanisms will be managed via the storage pad AMP:

  • Steel Loss of material due to general, pitting, and crevice corrosion Loss of material due to microbiological influenced corrosion
  • Concrete Loss of material due to freeze-thaw, aggressive chemical attack, corrosion of reinforcing steel, microbiological degradation, delayed ettringite formation 1, and salt scaling Cracking due to freeze-thaw, reaction with aggregates, differential settlement, aggressive chemical attack, corrosion of reinforcing steel, and delayed ettringite formation 1 Loss of strength due to reaction with aggregates, aggressive chemical attack, corrosion of reinforcing steel, leaching of calcium hydroxide, microbiological degradation, and delayed ettringite formation 1 Reduction of concrete pH due to aggressive chemical attack, leaching of calcium hydroxide, and microbiological degradation Loss of concrete/steel bond due to corrosion of reinforcing steel Increase in porosity and permeability due to leaching of calcium hydroxide and microbiological degradation 1

Delayed ettringite formation may be ruled out as a credible aging mechanism by the general licensee based on an ISFSI-specific evaluation.

Page 3-54

CoC 1027 Renewal Application Revision 0 / April 2020 3.7 Aging Management Review for Spent Fuel Assemblies Table 3-8 provides the detailed results of the AMR for the subcomponent parts of the SFAs. For each material group/environment combination and each part, the table lists the credible aging mechanisms and effects based on the information in Table 3-4.

The following aging effects/mechanisms for HBU fuel will be managed via the HBU Fuel AMP:

  • Zirconium-Based Alloy Loss of ductility due to hydride-induced embrittlement Changes in dimensions due to thermal creep Page 3-55

CoC 1027 Renewal Application Revision 0 / April 2020 3.8 Operating Experience Review Results - Aging Effects Identification The review of various sources of OE discussed in Appendix 3C did not identify any aging mechanisms and/or effects that were not already identified in NUREG-2214 [3-2]. In addition, no incidents were identified where aging effects lead to the loss of intended safety functions of TN-68 dry storage cask system SSCs. Therefore, it is concluded that the effects of aging will be managed adequately so that the SSC intended safety functions will be maintained during the PEO.

Page 3-56

CoC 1027 Renewal Application Revision 0 / April 2020 3.9 References 3-1 Nuclear Regulatory Commission, NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Revision 1, U.S. June 2016 (ADAMS Accession Number ML16179A148).

3-2 U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, July 2019.

3-3 TN-68 Dry Storage Cask Updated Final Safety Analysis Report, Revision 9, May 2018.

3-4 U.S. Nuclear Regulatory Commission, NUREG-1801, Generic Aging Lessons Learned (GALL) ReportFinal Report, Revision 2, December 2010.

3-5 U.S. Nuclear Regulatory Commission, CoC 1027 Appendix A, TN-68 Generic Technical Specifications, Amendment 1, October 30, 2007, Docket No. 72-1027.

3-6 U.S. Nuclear Regulatory Commission, Spent Fuel Project Office, Interim Staff Guidance 11,Cladding Considerations for the Transportation and Storage of Spent Fuel, Revision 3, November 17, 2003.

3-7 U.S. Nuclear Regulatory Commission, NUREG/CR-6428, Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds, May 1996.

3-8 U.S. Nuclear Regulatory Commission, NUREG/CR-6424, Report on Aging of Nuclear Power Plant Reinforced Concrete Structures, March 1996.

Page 3-57

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-1 Approved Fuel Designs Fuel Type Cladding Material Burnup Limit Source GE 7x7 Zircaloy 40 GWd/MTU [3-5]

GE 8x8 Zircaloy 60 GWd/MTU [3-5]

GE 9x9 Zircaloy 60 GWd/MTU [3-5]

GE 10x10 Zircaloy 60 GWd/MTU [3-5]

Page 3-58

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-2 Material Groupings Material Group Description Various carbon steels, alloy steels, and high-strength, low-alloy steels.

Steel Galvanized steel, aluminum-coated steel, and electroless nickel-plated steel, and are also included in the category of steel.

Stainless steel includes austenitic stainless steels and martensitic stainless Stainless Steel steel. Chrome-plated stainless steel is also included in the category of stainless steel.

Includes commercially pure aluminum 1100 and precipitation-hardened Aluminum alloys 6061 and 6063.

A mixture of hydraulic cement, aggregates, and water, with or without Concrete admixtures, fibers, or other cementitious materials.

Spent Fuel Assembly Cladding This grouping includes zircaloy cladding.

This group includes stainless steel, zirconium-based alloys, and nickel Spent Fuel Assembly Hardware alloys.

An aluminum alloy typically containing up to 4.5 weight percent boron.

It is used as a neutron poison material. The boron is incorporated in the aluminum matrix as discrete particles of AlB2 or TiB2 (for alloys also containing titanium). Aluminum alloys 1100, 6063, and 6351 have been Borated Aluminum used as base materials for boron additions.

This group also includes laminate composite neutron poison material consisting of a core of aluminum and boron-carbine powder sandwiched between sheets of aluminum, e.g., Boral.

This group is limited to the nickel alloy option for the liner of the Nickel Alloy metallic seals.

This group includes the polypropylene used in the top neutron shield and Polymers the borated polyester used in the radial neutron shield.

Page 3-59

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-3 Maximum TN-68 Cask Temperatures During Normal Storage Maximum Temperature Component Table 4B.3-1 of [3-3]

(°F)

Fuel Cladding 649 Basket 624 Basket Rails 402 Basket Shim 369 Inner Shell 340 Gamma Shield Shell 334 Radial Neutron Shield Tavg 300 Top Neutron Shield 227 Cask Outer Surface 277 Cask Lid Seal 227 Vent & Port Seals 227 Page 3-60

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-4 Summary of Potential Aging Mechanisms (4 Pages)

Material Credible Grouping Aging Mechanism Environments Section Aging Effect Air-Outdoor General Corrosion Embedded-in- 3.4.3.1.1 Loss of Material Concrete Air-Outdoor Pitting and Crevice Embedded-in- 3.4.3.1.2 Loss of Material Corrosion Concrete Galvanic Corrosion Air-Outdoor (5) 3.4.3.1.3 Loss of Material Microbiologically Influenced Embedded-in-3.4.3.1.4 Loss of Material Steel Corrosion Concrete (3)

Stress Corrosion Cracking None 3.4.3.1.5 None Creep None 3.4.3.1.6 None Fatigue None 3.4.3.1.7 None Thermal Aging None 3.4.3.1.8 None Radiation Embrittlement None 3.4.3.1.9 None (2)

Stress Relaxation Air-Outdoor 3.4.3.1.10 Loss of Preload Wear None 3.4.3.1.11 None General Corrosion None 3.4.3.2.1 None Pitting and Crevice Air-Outdoor 3.4.3.2.2 Loss of Material Corrosion Galvanic Corrosion Air-Outdoor (1) 3.4.3.2.3 Loss of Material Microbiologically Influenced None 3.4.3.2.4 None Corrosion Stainless Stress Corrosion Cracking None 3.4.3.2.5 None Steel Creep None 3.4.3.2.6 None Fatigue None 3.4.3.2.7 None Thermal Aging None 3.4.3.2.8 None Radiation Embrittlement None 3.4.3.2.9 None Stress Relaxation None 3.4.3.2.10 None Wear None 3.4.3.2.11 None General Corrosion Air-outdoor 3.4.3.3.1 Loss of Material Pitting and Crevice Air-outdoor 3.4.3.3.2 Loss of Material Corrosion Galvanic Corrosion Air-outdoor 3.4.3.3.3 Loss of Material Aluminum Microbiologically Influenced None 3.4.3.3.4 None Corrosion Creep None 3.4.3.3.5 None Fatigue None 3.4.3.3.6 None Page 3-61

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-4 Summary of Potential Aging Mechanisms (4 Pages)

Material Credible Grouping Aging Mechanism Environments Section Aging Effect Thermal Aging None 3.4.3.3.7 None Aluminum Radiation Embrittlement None 3.4.3.3.8 None General Corrosion None 3.4.3.4.1 None Pitting and Crevice None 3.4.3.4.2 None Corrosion Microbiologically Influenced Nickel None 3.4.3.4.3 None Corrosion Alloy Stress Corrosion Cracking None 3.4.3.4.4 None Fatigue None 3.4.3.4.5 None Radiation Embrittlement None 3.4.3.4.6 None Wear None 3.4.3.4.7 None Boron Depletion None 3.4.3.5.1 None Polymer Thermal Aging Fully encased 3.4.3.5.2 Shrinkage / Cracking Radiation Embrittlement Fully encased 3.4.3.5.3 Shrinkage / Cracking General Corrosion None 3.4.3.6.1 None Galvanic Corrosion None 3.4.3.6.2 None Boron Depletion None 3.4.3.6.3 None Borated Aluminum Creep None 3.4.3.6.4 None Thermal Aging None 3.4.3.6.5 None Radiation Embrittlement None 3.4.3.6.6 None Wet Corrosion and Blistering None 3.4.3.6.7 None Air-Outdoor Cracking Freeze-Thaw 3.4.3.7.1 Groundwater/Soil Loss of Material Creep None 3.4.3.7.2 None Concrete Air-Outdoor Cracking Reaction with Aggregates 3.4.3.7.3 Groundwater/Soil Loss of Strength Differential Settlement Groundwater/Soil 3.4.3.7.4 Cracking Page 3-62

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-4 Summary of Potential Aging Mechanisms (4 Pages)

Material Credible Grouping Aging Mechanism Environments Section Aging Effect Cracking Loss of Strength Air-Outdoor Aggressive Chemical Attack 3.4.3.7.5 Loss of Material Groundwater/Soil Reduction of Concrete pH Loss of Concrete/Steel Bond Corrosion of Reinforcing Air-Outdoor Loss of Material 3.4.3.7.6 Steel Groundwater/Soil Cracking Loss of Strength Shrinkage None 3.4.3.7.7 None Loss of Strength Air-Outdoor Increase in Porosity and Leaching of Calcium Groundwater/Soil 3.4.3.7.8 Permeability Hydroxide Reduction of Concrete pH Concrete Radiation Damage None 3.4.3.7.9 None Fatigue None 3.4.3.7.10 None Dehydration at High None 3.4.3.7.11 None Temperature Loss of Strength Loss of Material Microbiological Degradation Groundwater/Soil 3.4.3.7.12 Increase in Porosity and Permeability Reduction of Concrete pH Loss of Material Air-Outdoor (4)

Delayed Ettringite Formation 3.4.3.7.13 Loss of Strength Groundwater/Soil (4)

Cracking Air-Outdoor Salt Scaling 3.4.3.7.14 Loss of Material Groundwater/Soil Hardening None 3.4.3.7.15 None Page 3-63

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-4 Summary of Potential Aging Mechanisms (4 Pages)

Material Credible Grouping Aging Mechanism Environments Section Aging Effect Hydride-Induced Helium(6) 3.4.3.8.1 Loss of Ductility Embrittlement Delayed Hydride Cracking None 3.4.3.8.2 None (6)

Thermal Creep Helium 3.4.3.8.3 Change in Dimensions Low Temperature Creep None 3.4.3.8.4 None Spent Fuel Mechanical Overload None 3.4.3.8.5 None Assembly Oxidation None 3.4.3.8.6 None Cladding Pitting Corrosion None 3.4.3.8.7 None Galvanic Corrosion None 3.4.3.8.8 None Stress Corrosion Cracking None 3.4.3.8.9 None Radiation Embrittlement None 3.4.3.8.10 None Fatigue None 3.4.3.8.11 None Creep None 3.4.3.9.1 None Hydriding None 3.4.3.9.2 None Spent Fuel General Corrosion None 3.4.3.9.3 None Assembly Hardware Stress Corrosion Cracking None 3.4.3.9.4 None Radiation Embrittlement None 3.4.3.9.5 None Fatigue None 3.4.3.9.6 None

1. Only for stainless steel in contact with a graphite lubricant.
2. Only for bolt subcomponents.
3. Only if the concrete is exposed to a groundwater/soil environment.
4. DEF is credible for the storage pad only, unless ruled out by the general licensee based on ISFSI-specific evaluation.
5. Only for steel in contact with more noble materials such as stainless steel or a graphite lubricant.
6. Only for HBU fuel.

Page 3-64

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-5 Aging Management Review for TN-68 Cask (4 Pages)

UFSAR Drawing Intended Safety Credible Aging Aging Management Component No. Item No. Subcomponent Parts Function (1) Material Group Environment (2) Mechanism Aging Effect Activity General Corrosion Loss of Material AMP Air-Outdoor (E) Pitting and Crevice Loss of Material AMP TN-68 972-70-2 1 Gamma Shield SH, TH, SR, RT Steel Corrosion Fully Encased (E) None None None Fully Encased (I) None None None General Corrosion Loss of Material AMP Pitting and Crevice Lid Air-outdoor(E) Loss of Material AMP Steel Corrosion (including stainless steel TN-68 972-70-2 2 CO, SH, TH, SR, Galvanic Corrosion Loss of Material AMP weld overlay at seal location) Fully Encased (I) None None None Stainless Steel Fully Encased None None None Fully Encased (E) None None None TN-68 972-70-2 3 Inner Containment CO, SH, TH, SR, Steel Helium (I) None None None Fully Encased (I) None None None General Corrosion Loss of Material AMP TN-68 972-70-2 4 Bottom SH, TH, SR, RT Steel Air-Outdoor (E) Pitting and Crevice Loss of Material AMP Corrosion Fully Encased (E) None None None TN-68 972-70-2 5 Bottom Containment CO, SH, TH, SR, Steel Helium (I) None None None Fully Encased (I) None None None TN-68 972-70-2 6 Upper Trunnion SH, SR, RT Stainless Steel Pitting and Crevice Air-Outdoor (E) Loss of Material AMP Corrosion Fully Encased (I) None None None General Corrosion Loss of Material AMP TN-68 972-70-2 7 Lower Trunnion SH, SR, RT Steel Air-Outdoor (E) Pitting and Crevice Loss of Material AMP Corrosion Helium (I) None None None TN-68 972-70-2 8 Shield Plate SH, TH, SR Steel Fully Encased (E) None None None Thermal Aging Shrinkage / Cracking AMP TN-68 972-70-2 9 Radial Neutron Shield SH, TH Polymer Fully Encased Radiation Shrinkage / Cracking AMP Embrittlement Fully Encased (I) None None None General Corrosion Loss of Material AMP TN-68 972-70-2 10 Outer Shell SH, TH, SR Steel Air-Outdoor (E) Pitting and Crevice Loss of Material AMP Corrosion Page 3-65

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-5 Aging Management Review for TN-68 Cask (4 Pages)

UFSAR Drawing Intended Safety Credible Aging Aging Management Component No. Item No. Subcomponent Parts Function (1) Material Group Environment (2) Mechanism Aging Effect Activity General Corrosion Loss of Material AMP Pitting and Crevice TN-68 972-70-2 11 Protective Cover SH, TH Steel Air-Outdoor Loss of Material AMP Corrosion Galvanic Corrosion Loss of Material AMP Thermal Aging Shrinkage / Cracking AMP (4)

TN-68 972-70-2 12 Top Neutron Shield SH, TH Polymer Fully Encased Radiation Shrinkage / Cracking AMP Embrittlement TN-68 972-70-2 13 Radial N-Shield Box SH, TH Aluminum Fully Encased None None None General Corrosion Loss of Material AMP Pitting and Crevice Loss of Material AMP TN-68 972-70-2 14 Lid Bolt CO, SH, TH, SR Steel Air-Outdoor Corrosion Galvanic Corrosion Loss of Material AMP Stress Relaxation Loss of Preload AMP Nickel alloy Helium None None None Stainless Steel Helium None None None General Corrosion Loss of Material AMP TN-68 972-70-2 16 Lid Seal CO Pitting and Crevice Aluminum Air-Outdoor Loss of Material AMP Corrosion Galvanic Corrosion Loss of Material AMP Helium (I) None None None TN-68 972-70-2 18 Overpressure Port Cover CO Stainless Steel Pitting and Crevice Air-Outdoor (E) Loss of Material AMP Corrosion Nickel alloy Helium None None None Stainless Steel Helium None None None Overpressure Port Cover General Corrosion Loss of Material AMP TN-68 972-70-2 19 CO Seal Pitting and Crevice Aluminum Air-Outdoor Loss of Material AMP Corrosion Galvanic Corrosion Loss of Material AMP Helium (I) None None None TN-68 972-70-2 22 Drain Port Cover CO, SH, TH, SR Stainless Steel Pitting and Crevice Air-Outdoor (E) Loss of Material AMP Corrosion Helium (I) None None None TN-68 972-70-2 23 Vent Port Cover CO, SH, TH, SR Stainless Steel Pitting and Crevice Air-Outdoor (E) Loss of Material AMP Corrosion Page 3-66

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-5 Aging Management Review for TN-68 Cask (4 Pages)

UFSAR Drawing Intended Safety Credible Aging Aging Management Component No. Item No. Subcomponent Parts Function (1) Material Group Environment (2) Mechanism Aging Effect Activity Nickel alloy Helium None None None Stainless Steel Helium None None None Vent & Drain Port Cover General Corrosion Loss of Material AMP TN-68 972-70-2 24 CO Seal Pitting and Crevice Aluminum Air-Outdoor Loss of Material AMP Corrosion Galvanic Corrosion Loss of Material AMP General Corrosion Loss of Material AMP Pitting and Crevice Vent & Drain Port Cover Loss of Material AMP TN-68 972-70-2 25 CO, SR Steel Air-Outdoor Corrosion Bolts (SOC HD Cap)

Galvanic Corrosion Loss of Material AMP Stress Relaxation Loss of Preload AMP General Corrosion Loss of Material AMP Pitting and Crevice Overpressure Port Cover Loss of Material AMP TN-68 972-70-2 26 CO, SR Steel Air-Outdoor Corrosion Bolts (SOC HD Cap)

Galvanic Corrosion Loss of Material AMP Stress Relaxation Loss of Preload AMP TN-68 972-70-2 28 Basket Rail, Type 1 CR, TH, SR Aluminum Helium None None None TN-68 972-70-2 29 Basket Rail, Type 1 CR, TH, SR Aluminum Helium None None None TN-68 972-70-2 30 Basket Shim CR, TH, SR Stainless Steel Helium None None None TN-68 972-70-2 31 Basket Shim TH Aluminum Helium None None None TN-68 972-70-2 32 Fuel Compartment SH, CR, TH, SR Stainless Steel Helium None None None TN-68 972-70-2 33a Poison Plate SH, CR, TH Borated Aluminum Helium None None None TN-68 972-70-2 33b Aluminum Plate SH, CR, TH Aluminum Helium None None None TN-68 972-70-2 34 Structural Plates CR, TH, SR Stainless Steel Helium None None None Helium (I) None None None Fully Encased (E) None None None General Corrosion Loss of Material AMP Steel Pitting and Crevice Flange (including Stainless Air-Outdoor (E) Loss of Material AMP TN-68 972-70-2 35 CO, SH, TH, SR Corrosion Steel Weld Overlay)

Galvanic Corrosion Loss of Material AMP Fully Encased (I) None None None Stainless Steel Pitting and Crevice Air-Outdoor (E) Loss of Material AMP Corrosion TN-68 972-70-2 36 Shim SH Steel Fully Encased None None None Page 3-67

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-5 Aging Management Review for TN-68 Cask (4 Pages)

UFSAR Drawing Intended Safety Credible Aging Aging Management Component No. Item No. Subcomponent Parts Function (1) Material Group Environment (2) Mechanism Aging Effect Activity General Corrosion Loss of Material AMP Pitting and Crevice Loss of Material AMP TN-68 972-70-2 37 Trunnion Bolt SR, RT Steel Air-Outdoor Corrosion Galvanic Corrosion Loss of Material AMP Stress Relaxation Loss of Preload AMP TN-68 972-70-2 39 Basket Hold down SH, CR Stainless Steel Helium None None None General Corrosion Loss of Material AMP Pitting and Crevice TN-68 972-70-2 45 Threaded Insert CO, SR Steel Air-Outdoor Loss of Material AMP Corrosion Galvanic Corrosion Loss of Material AMP Pitting and Crevice Loss of Material AMP TN-68 972-70-2 47 SOC Head Cap Screw SH, SR Stainless Steel Air-Outdoor Corrosion Galvanic Corrosion Loss of Material AMP General Corrosion Loss of Material AMP TN-68 972-70-2 48 Shield Ring SH Steel Air-Outdoor Pitting and Crevice Loss of Material AMP Corrosion Fuel Compartment TN-68 972-70-7 52 CR Stainless Steel Helium None None None Extension TN-68 972-70-7 53 End Cap, Top CR Stainless Steel Helium None None None TN-68 972-70-7 54 End Cap Bottom CR Stainless Steel Helium None None None

1. The intended safety functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR), Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT)
2. If the subcomponent has an internal and external surface exposed to different environments, (I) refers to an internal (or towards the interior of the TN-68 cask) environment and (E) refers to an external (or towards the exterior of the TN-68 cask) environment.
3. The metallic seals consist of an inner spring, a lining, and a jacket. The spring is [ ] (a nickel alloy) or an equivalent material. The lining is stainless steel or nickel alloy. The jacket is made of aluminum.
4. Although it is not shown on the UFSAR Drawings, the top neutron shield polypropylene material is encased in carbon steel.

Page 3-68

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-6 Aging Management Review Results Summary TN-68 Cask (6 Pages)

Intended Aging Subcomponent UFSAR Drawing Safety Material Credible Aging Management Parts (Part #s) Function(s) (1) Group Environment (2) Mechanism Aging Effect Activity Loss of General Corrosion AMP 972-70-2 SH, TH, SR, Material Gamma Shield Steel Air-Outdoor (E)

(1) RT Pitting and Crevice Loss of AMP Corrosion Material Loss of Lid General Corrosion AMP Material (including 972-70-2 CO, SH, TH, Pitting and Crevice Loss of stainless steel Steel Air-outdoor(E) AMP (2) SR, Corrosion Material weld overlay at seal location) Loss of Galvanic Corrosion AMP Material Loss of General Corrosion AMP 972-70-2 SH, TH, SR, Material Bottom Steel Air-Outdoor (E)

(4) RT Pitting and Crevice Loss of AMP Corrosion Material 972-70-2 Stainless Pitting and Crevice Loss of Upper Trunnion SH, SR, RT Air-Outdoor (E) AMP (6) Steel Corrosion Material Loss of General Corrosion AMP 972-70-2 Material Lower Trunnion SH, SR, RT Steel Air-Outdoor (E)

(7) Pitting and Crevice Loss of AMP Corrosion Material Shrinkage /

Thermal Aging AMP Radial Neutron 972-70-2 Cracking SH, TH Polymer Fully Encased Shield (9) Radiation Shrinkage /

AMP Embrittlement Cracking Page 3-69

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-6 Aging Management Review Results Summary TN-68 Cask (6 Pages)

Intended Aging Subcomponent UFSAR Drawing Safety Material Credible Aging Management Parts (Part #s) Function(s) (1) Group Environment (2) Mechanism Aging Effect Activity Loss of General Corrosion AMP 972-70-2 Material Outer Shell SH, TH, SR Steel Air-Outdoor (E)

(10) Pitting and Crevice Loss of AMP Corrosion Material Loss of General Corrosion AMP Material 972-70-2 Pitting and Crevice Loss of Protective Cover SH, TH Steel Air-Outdoor AMP (11) Corrosion Material Loss of Galvanic Corrosion AMP Material Shrinkage /

Thermal Aging AMP Top Neutron 972-70-2 Cracking SH, TH Polymer Fully Encased Shield (12) Radiation Shrinkage /

AMP Embrittlement Cracking Loss of General Corrosion AMP Material Pitting and Crevice Loss of AMP 972-70-2 CO, SH, TH, Corrosion Material Lid Bolt Steel Air-Outdoor (14) SR Loss of Galvanic Corrosion AMP Material Loss of Stress Relaxation AMP Preload Page 3-70

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-6 Aging Management Review Results Summary TN-68 Cask (6 Pages)

Intended Aging Subcomponent UFSAR Drawing Safety Material Credible Aging Management Parts (Part #s) Function(s) (1) Group Environment (2) Mechanism Aging Effect Activity Loss of General Corrosion AMP Material 972-70-2 Pitting and Crevice Loss of Lid Seal CO Aluminum Air-Outdoor AMP (16) Corrosion Material Loss of Galvanic Corrosion AMP Material Overpressure Port 972-70-2 Stainless Pitting and Crevice Loss of CO Air-Outdoor (E) AMP Cover (18) Steel Corrosion Material Loss of General Corrosion AMP Material Overpressure Port 972-70-2 Pitting and Crevice Loss of CO Aluminum Air-Outdoor AMP Cover Seal (19) Corrosion Material Loss of Galvanic Corrosion AMP Material 972-70-2 CO, SH, TH, Stainless Pitting and Crevice Loss of Drain Port Cover Air-Outdoor (E) AMP (22) SR Steel Corrosion Material 972-70-2 CO, SH, TH, Stainless Pitting and Crevice Loss of Vent Port Cover Air-Outdoor (E) AMP (23) SR Steel Corrosion Material Loss of General Corrosion AMP Material Vent & Drain 972-70-2 Pitting and Crevice Loss of CO Aluminum Air-Outdoor (E) AMP Port Cover Seal (24) Corrosion Material Loss of Galvanic Corrosion AMP Material Page 3-71

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-6 Aging Management Review Results Summary TN-68 Cask (6 Pages)

Intended Aging Subcomponent UFSAR Drawing Safety Material Credible Aging Management Parts (Part #s) Function(s) (1) Group Environment (2) Mechanism Aging Effect Activity Loss of General Corrosion AMP Material Pitting and Crevice Loss of Vent & Drain AMP 972-70-2 Corrosion Material Port Cover Bolts CO, SR Steel Air-Outdoor (25) Loss of (SOC HD Cap) Galvanic Corrosion AMP Material Loss of Stress Relaxation AMP Preload Loss of General Corrosion AMP Material Pitting and Crevice Loss of Overpressure Port AMP 972-70-2 Corrosion Material Cover Bolts (SOC CO, SR Steel Air-Outdoor (26) Loss of HD Cap) Galvanic Corrosion AMP Material Loss of Stress Relaxation AMP Preload Loss of General Corrosion AMP Material Pitting and Crevice Loss of Steel Air-Outdoor (E) AMP 972-70-2 CO, SH, TH, Corrosion Material Flange (35) SR Loss of Galvanic Corrosion AMP Material Stainless Pitting and Crevice Loss of Air-Outdoor (E) AMP Steel Corrosion Material Page 3-72

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-6 Aging Management Review Results Summary TN-68 Cask (6 Pages)

Intended Aging Subcomponent UFSAR Drawing Safety Material Credible Aging Management Parts (Part #s) Function(s) (1) Group Environment (2) Mechanism Aging Effect Activity Loss of General Corrosion AMP Material Pitting and Crevice Loss of AMP 972-70-2 Corrosion Material Trunnion Bolt SR, RT Steel Air-Outdoor (37) Loss of Galvanic Corrosion AMP Material Loss of Stress Relaxation AMP Preload Loss of General Corrosion AMP Material 972-70-2 Pitting and Crevice Loss of Threaded Inserts CO, SR Steel Air-Outdoor AMP (45) Corrosion Material Loss of Galvanic Corrosion AMP Material Pitting and Crevice Loss of AMP SOC Head Cap 972-70-2 Stainless Corrosion Material SH, SR Air-Outdoor Screw (47) Steel Loss of Galvanic Corrosion AMP Material Loss of General Corrosion AMP 972-70-2 Material Shield Ring SH Steel Air-Outdoor (48) Pitting and Crevice Loss of AMP Corrosion Material

1. The intended safety functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR), Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT)
2. If the subcomponent has an internal and external surface exposed to different environments, (I) refers to an internal (or towards the interior of the TN-68 cask) environment and (E) refers to an external (or towards the exterior of the TN-68 cask) environment.

Page 3-73

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-7 Aging Management Review for Storage pad (3 Pages)

Intended Aging Subcomponent Safety Material Management Parts Function(s) (1) Group Environment Credible Aging Mechanism Aging Effect Activity General Corrosion Loss of Material Storage pad AMP Pitting and Crevice Embedded-in- Loss of Material Storage pad AMP Steel Corrosion Concrete Microbiologically Influenced Loss of Material Storage Pad AMP Corrosion Cracking Storage pad AMP Freeze-Thaw Loss of Material Storage pad AMP Cracking Storage pad AMP Reaction with Aggregates Loss of Strength Storage pad AMP Cracking Storage pad AMP Storage Pad TH Loss of Strength Storage pad AMP Aggressive Chemical Attack Loss of Material Storage pad AMP Reduction of Concrete pH Storage pad AMP Concrete Air-Outdoor Loss of Concrete/Steel Bond Storage pad AMP Corrosion of Reinforcing Loss of Material Storage pad AMP Steel Cracking Storage pad AMP Loss of Strength Storage pad AMP Loss of Strength Storage pad AMP Leaching of Calcium Increase in Porosity and Storage pad AMP Hydroxide Permeability Reduction of Concrete pH Storage pad AMP Page 3-74

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-7 Aging Management Review for Storage pad (3 Pages)

Intended Aging Subcomponent Safety Material Management Parts Function(s) (1) Group Environment Credible Aging Mechanism Aging Effect Activity Loss of Material Storage pad AMP Delayed Ettringite Loss of Strength Storage pad AMP Air-Outdoor Formation (2)

Cracking Storage pad AMP Salt Scaling Loss of Material Storage pad AMP Cracking Storage pad AMP Freeze-Thaw Loss of Material Storage pad AMP Cracking Storage pad AMP Reaction with Aggregates Loss of Strength Storage pad AMP Differential Settlement Cracking Storage pad AMP Cracking Storage pad AMP Storage Pad TH Concrete Loss of Strength Storage pad AMP Aggressive Chemical Attack Groundwater/ Loss of Material Storage pad AMP Soil Reduction of Concrete pH Storage pad AMP Loss of Concrete/Steel Bond Storage pad AMP Corrosion of Reinforcing Loss of Material Storage pad AMP Steel Cracking Storage pad AMP Loss of Strength Storage pad AMP Loss of Strength Storage pad AMP Leaching of Calcium Increase in Porosity and Storage pad AMP Hydroxide Permeability Reduction of Concrete pH Storage pad AMP Page 3-75

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-7 Aging Management Review for Storage pad (3 Pages)

Intended Aging Subcomponent Safety Material Management Parts Function(s) (1) Group Environment Credible Aging Mechanism Aging Effect Activity Loss of Strength Storage pad AMP Loss of Material Storage pad AMP Microbiological Degradation Increase in Porosity and Storage pad AMP Permeability Groundwater/ Reduction of Concrete pH Storage pad AMP Storage Pad TH Concrete Soil Loss of Material Storage pad AMP Delayed Ettringite Loss of Strength Storage pad AMP Formation (2)

Cracking Storage pad AMP Salt Scaling Loss of Material Storage pad AMP

1. The intended safety functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR), Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT).
2. Delayed Ettringite Formation may be ruled out as a credible aging mechanism by the general licensee based on an ISFSI-specific evaluation.

Page 3-76

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3-8 Aging Management Review for Spent Fuel Assemblies (2 pages)

Intended Aging Safety Material of Credible Aging Management Subcomponent Function (1) Construction Material Group Environment (2) Mechanism Aging Effect Activity Hydride-Induced Loss of HBU Fuel Embrittlement(3) Ductility(3) AMP (I) Helium Changes in HBU Fuel Thermal Creep(3)

Fuel Cladding and CO, CR, SR, Spent Fuel Assembly Dimensions(3) AMP Zircaloy End Plugs TH Cladding Hydride-Induced Loss of HBU Fuel Embrittlement(3) Ductility(3) AMP (E) Helium Changes in HBU Fuel Thermal Creep(3)

Dimensions(3) AMP Spent Fuel Assembly Inconel Helium None None No Hardware Spacer Grid CR, SR Assemblies Spent Fuel Assembly Zircaloy Helium None None No Hardware Upper End Spent Fuel Assembly Stainless Steel Helium None None No Fitting/Nozzle Hardware SR (and Related Spent Fuel Assembly subcomponents) Inconel Helium None None No Hardware Lower End Spent Fuel Assembly Stainless Steel Helium None None No Fitting/Nozzle Hardware SR (and Related Spent Fuel Assembly subcomponents) Inconel Helium None None No Hardware Spent Fuel Assembly Water Channels CR, SR Zircaloy Helium None None No Hardware Spent Fuel Assembly Fuel Channel CR, TH Zircaloy Helium None None No Hardware

1. The intended safety functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR), Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT)

Page 3-77

CoC 1027 Renewal Application Revision 0 / April 2020

2. If the subcomponent has an internal and external surface exposed to different environments, (I) refers to the internal environment of the fuel pin and (E) refers to an external environment of the fuel pin.
3. For HBU fuel only.

Page 3-78

CoC 1027 Renewal Application Revision 0 / April 2020 APPENDIX 3A TLAA IDENTIFICATION AND DISPOSITION CONTENTS 3A.1 Introduction ................................................................................................................. 3A-1 3A.2 Methodology for Identification and Disposition of TLAAs .................................... 3A-2 3A.2.1 Identification of TLAAs ................................................................................ 3A-2 3A.2.2 Disposition of Identified TLAAs ................................................................... 3A-2 3A.3 Identified TLAAs ........................................................................................................ 3A-4 3A.4 Disposition of Identified TLAAs ................................................................................ 3A-5 3A.4.1 End of Life Cavity Pressure ........................................................................... 3A-5 3A.5 Summary of TLAA Identification and Disposition.................................................. 3A-6 3A.6 Summary of Updated End of Life Cavity Pressure Analysis.................................. 3A-7 3A.7 References (Appendix A, TLAA Identification and Disposition) ........................... 3A-8 LIST OF TABLES Table 3A-1 Identified TLAAs .............................................................................................. 3A-9 Page 3A-i

CoC 1027 Renewal Application Revision 0 / April 2020 3A.1 Introduction Per Code of Federal Regulations (CFR) 10 CFR 72.240(c)(2) [3A-1], a certificate of compliance (CoC) renewal application must include time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to nuclear safety will continue to perform their intended safety functions for the requested period of extended operation (PEO). For an analysis to be considered a TLAA, it must meet the six selection criteria defined in 10 CFR 72.3 [3A-1].

This appendix describes the process used to identify and disposition TLAAs. It also summarizes the TLAAs that were updated to demonstrate that they have been projected to the end of the PEO.

Note that the initial license for CoC No. 1027 was for a 20-year period and that the license renewal application will request an additional 40 years. Therefore, the total PEO will be 60 years.

Page 3A-1

CoC 1027 Renewal Application Revision 0 / April 2020 3A.2 Methodology for Identification and Disposition of TLAAs 3A.2.1 Identification of TLAAs TLAAs are calculations or analyses used to demonstrate that in-scope SSCs will maintain their intended safety function throughout an explicitly stated PEO. To be considered a TLAA, the calculation/analysis must meet all six of the following criteria as defined in 10 CFR 72.3 [3A-1]:

1. Involves SSCs important to safety within the scope of the spent fuel storage certificate renewal, as delineated in Subpart L of 10 CFR Part 72 [3A-1];
2. Considers the effects of aging;
3. Involves time-limited assumptions defined by the current operating term;
4. Was determined to be relevant by the certificate holder in making a safety determination;
5. Involves conclusions or basis of conclusions related to capability of the SSCs to perform their intended safety functions; and
6. Is contained or incorporated by reference in the design bases.

TLAAs were identified via a review of the updated final safety analysis report (UFSAR) [3A-3] and the TN-68 design calculations.

The review of the UFSAR involved performing word searches for key words (i.e.,

year, yr, hour, hr, life, cycle, aging, fatigue).

For each hit of a key word, the context of the key word was then reviewed to determine if a time-limited assumption used in an analysis was involved, i.e., TLAA Criterion 3. If it was determined that a time limiting-assumption was not involved, the review stopped at that point. If a time-limited assumption was involved, then the context was reviewed until another criterion was not met. If all criteria were met, then the item involved a TLAA.

The review of the design basis calculations followed a similar process (i.e., first determining if a time-limited assumption was involved and if one was, then continue review until another criterion was not meet.) If all the criteria were met, then the calculation involved a TLAA.

3A.2.2 Disposition of Identified TLAAs The identified TLAAs were dispositioned using one of the following three methods listed in Section 3.5.1 of NUREG-1927 [3A-2]:

1. Demonstrate the existing analysis remains valid for the PEO, has already considered the requested PEO, and concludes that the SSC will continue to perform its intended safety function through the end of the requested PEO; Page 3A-2

CoC 1027 Renewal Application Revision 0 / April 2020

2. Revise or update the existing analysis to demonstrate that it has been projected to the end of the requested PEO and concludes that the SSC will continue to perform its intended safety function through the end of the requested PEO; and
3. Manage the effects of aging on the SSC for the requested PEO through an aging management program (AMP).

Page 3A-3

CoC 1027 Renewal Application Revision 0 / April 2020 3A.3 Identified TLAAs Table 3A-1 provides a summary of the identified TLAAs.

Page 3A-4

CoC 1027 Renewal Application Revision 0 / April 2020 3A.4 Disposition of Identified TLAAs 3A.4.1 End of Life Cavity Pressure USFAR [3A-3] Section 2.2.5.3.3 contains a design criteria that 1 atmosphere of pressure must exist in the cavity on the coldest day at the end of life. The purpose of pressurizing the cavity above atmospheric pressure is to prevent in-leakage of air.

While the design of the overpressure monitoring system ensures there will be no out-leakage past the seals during normal operations, the cavity pressure decreases over time as the decay heat decreases. The end of life cavity pressure analysis described in USFAR [3A-3] Section 7.2.2 assumes a cavity gas temperature corresponding to the end of a 40-year storage period.

Therefore, the end of life cavity pressure analysis must be revised or updated to demonstrate that the cavity pressure will be above one atmosphere at the end of the PEO, i.e., a total of 60 years.

Page 3A-5

CoC 1027 Renewal Application Revision 0 / April 2020 3A.5 Summary of TLAA Identification and Disposition The review of the approved design basis for CoC 1027 identified the following TLAA:

  • ensure cavity pressure remains above one atmosphere on the coldest day at the storage period The following TLAA must be revised or updated (see Section 3A.6 for updated analysis) to demonstrate that it has been projected to the end of the PEO and concludes that the SSC will continue to perform its intended function through the end of the PEO:
  • ensure cavity pressure remains above one atmosphere on the coldest day at the storage period Page 3A-6

CoC 1027 Renewal Application Revision 0 / April 2020 3A.6 Summary of Updated End of Life Cavity Pressure Analysis This section summarizes the calculation that demonstrates that the cavity pressure of a TN-68 cask remains above atmospheric pressure on the coldest day at the end of the PEO.

The ideal gas law governs the change in cavity pressure as the temperature inside the cavity decreases so that P2= P1

  • T2 / T1 where, P1 = is the pressure at the beginning of storage; T1 = is the temperature at the beginning of storage; P2 = is the pressure at the end of storage; and T2 = is the temperature at the end of storage.

The backfill pressure of a TN-68 cavity is controlled by Technical Specification 3.1.2

[3A-4] Cask Helium Backfill Pressure, i.e. 2.0 atm (+0/-10%). Therefore, the calculation conservatively assumes an initial pressure of 1.8 atm.

From Section 7.2.2.1 of the UFSAR [3A-3], the cavity gas temperature at the time the backfilling of the cask is complete is 350 °F (810 °R).

The average cavity gas temperature at the end of 60 years of storage was determined to be 47.5 °F (507 °R) based on a decay heat value of 8.42 kW and an ambient temperature of -20 °F.

Therefore, the pressure at the end of storage on the coldest day is P2 = 1.8 atm * (507 °R / 810 °R)

P2 = 1.127 atm Since this pressure is greater than 1 atm, the cavity gas pressure within a TN-68 cask will remain above atmospheric pressure on the coldest day at the end of the storage period.

Page 3A-7

CoC 1027 Renewal Application Revision 0 / April 2020 3A.7 References (Appendix A, TLAA Identification and Disposition) 3A-1 U.S. Nuclear Regulatory Commission, 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste, Code of Federal Regulations.

3A-2 U.S. Nuclear Regulatory Commission, NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Revision 1, June 2016.

3A-3 TN-68 Dry Storage Cask Updated Final Safety Analysis Report, Revision 9, May 2018.

3A-4 U.S. Nuclear Regulatory Commission, CoC 1027 Appendix A, TN-68 Generic Technical Specifications, Amendment 1, October 30, 2017, Docket No. 72-1027.

Page 3A-8

CoC 1027 Renewal Application Revision 0 / April 2020 Table 3A-1 Identified TLAAs SSC Involved Aging Effect Involved UFSAR Section Comment All subcomponents Ensures cask cavity within the Loss of material due to corrosion(1) UFSAR Sect 7.2.2 remains above 1 atm at confinement end of life barrier (1)

While ensuring that the cask cavity pressure is greater than 1 atmosphere is not a direct aging effect, it is intended to ensure that air does not leak into the cask cavity. Air in-leakage could lead to loss of material due to corrosion, which is an aging effect.

Page 3A-9

CoC 1027 Renewal Application Revision 0 / April 2020 APPENDIX 3B SUPPLEMENTAL EVALUATIONS CONTENTS 3B.1 Introduction ..................................................................................................................3B-1 LIST OF TABLES None Page 3B-i

CoC 1027 Renewal Application Revision 0 / April 2020 3B.1 Introduction No supplemental evaluations were performed to support the AMR and/or an element in an AMP associated with the renewal of CoC 1027.

Page 3B-1

CoC 1027 Renewal Application Revision 0 / April 2020 APPENDIX 3C Operating Experience Review CONTENTS 3C.1 Introduction ................................................................................................................. 3C-1 3C.2 Operating Experience Review Approach ................................................................. 3C-2 3C.3 Internal and Industry Condition Reports ................................................................ 3C-3 3C.3.1 Review of Internal Condition Reports ........................................................... 3C-3 3C.3.2 Review of CoC Users Issue Reports .............................................................. 3C-3 3C.3.3 Review of Industry Condition Reports .......................................................... 3C-3 3C.4 Relevant International and Non-Nuclear Operating Experience ........................... 3C-4 3C.5 Previous ISFSI Inspection Results ............................................................................ 3C-5 3C.5.1 Prairie Island ISFSI Pre-Application Inspection............................................ 3C-5 3C.5.2 North Anna ISFSI Pre-Application Inspection .............................................. 3C-6 3C.6 Licensee Event Reports .............................................................................................. 3C-8 3C.7 Vendor-Issued Safety Bulletins.................................................................................. 3C-9 3C.8 NRC Generic Communications ............................................................................... 3C-10 3C.8.1 NRC Information Notices ............................................................................ 3C-10 3C.8.2 NRC Bulletins .............................................................................................. 3C-10 3C.9 Updated Consensus Codes, Standards, or Guides ................................................. 3C-12 3C.9.1 Aging Management Reviews ....................................................................... 3C-12 3C.9.2 Aging Management Programs ..................................................................... 3C-12 3C.10 Applicable Industry Initiatives ................................................................................ 3C-13 3C.10.1 High Burnup Fuel Demonstration Project ................................................... 3C-13 3C.10.2 Chloride-Induced Stress Corrosion Cracking .............................................. 3C-13 3C.11 Conclusion ................................................................................................................. 3C-14 3C.12 References .................................................................................................................. 3C-15 LIST OF TABLES Table 3C-1 TN Americas Internal CARs ............................................................................3C-16 Table 3C-2 Issue Reports from Peach Bottom Nuclear Station..........................................3C-18 Table 3C-3 TNUG Technical Bulletins ..............................................................................3C-20 Page 3C-i

CoC 1027 Renewal Application Revision 0 / April 2020 3C.1 Introduction This Appendix summarizes the operating experience (OE) review conducted for the renewal of Certificate of Compliance (CoC) No. 1027. The OE review is intended to provide objective evidence to support (or refute) the conclusion that the effects of aging will be managed adequately so that the in-scope structures, systems, and components (SSCs) intended safety functions will be maintained during the period of extended operation (PEO).

Page 3C-1

CoC 1027 Renewal Application Revision 0 / April 2020 3C.2 Operating Experience Review Approach This is a document review of the sources of OE listed below looking for aging-related (versus event-driven) degradation.

The sources of potential OE reviewed are:

  • Internal and industry wide condition reports
  • Relevant international and non-nuclear OE
  • Previous independent spent fuel storage installation (ISFSI) inspection results
  • Licensee event reports
  • Vendor-issued safety bulletins
  • U.S. Nuclear Regulatory Commission (NRC) Generic Communications
  • Updated consensus codes, standards, or guides
  • Applicable industry-initiatives Page 3C-2

CoC 1027 Renewal Application Revision 0 / April 2020 3C.3 Internal and Industry Condition Reports 3C.3.1 Review of Internal Condition Reports Table 3C-1 provides a summary of the TN Americas LLC corrective action reports (CARs) related to bolted casks components (e.g., TN-32, TN-40, and TN-68) since 2002. As shown in Table 3C-1, most conditions were either event-driven or involve subcomponents that are not in-scope for the renewal of CoC 1027. Those conditions that were aging-related involved aging mechanisms/effects already identified as requiring management.

3C.3.2 Review of CoC Users Issue Reports Currently, there is a single ISFSI storing spent fuel using CoC 1027, the Peach Bottom Nuclear Station. Table 3C-2 provides a summary of the Peach Bottom ISFSI issue reports. As shown in Table 3C-2, most conditions were either event-driven or involve subcomponents that are not in-scope for the renewal of CoC 1027. Those conditions that were aging related involved aging mechanisms/effects already identified as requiring management.

3C.3.3 Review of Industry Condition Reports A review of the Aging Management INPO Database (AMID) did not identify any reported OE associated with the TN-68 dry storage cask system. Also, AMID did not contain any reported OE for the similar TN-32 and TN-40 storage systems.

Page 3C-3

CoC 1027 Renewal Application Revision 0 / April 2020 3C.4 Relevant International and Non-Nuclear Operating Experience The aging management review (AMR) conducted for the renewal of CoC 1027 is based on the aging mechanisms/effects described in NUREG-2214 [3C-8]. Since the information in this report is based, in part, on international and nonnuclear OE, the AMR is also based in part on international and nonnuclear OE.

Page 3C-4

CoC 1027 Renewal Application Revision 0 / April 2020 3C.5 Previous ISFSI Inspection Results 3C.5.1 Prairie Island ISFSI Pre-Application Inspection In the summer of 2011, Northern States Power Company of Minnesota (NSPM) conducted a license renewal pre-application inspection [3C-6] of TN-40 casks at their Prairie Island Nuclear Generating Power plant ISFSI. In addition to the accessible exterior surfaces of a cask (Cask 01), the inspection also included visual inspections on the bottom of a cask and the area underneath the protective cover. Additional inspections were performed of the area underneath the protective cover of a second cask (Cask 13). The results of the pre-application inspection are summarized in [3C-6].

The results of the cask bottom inspection revealed that approximately 25% of the protective coating on the bottom of Cask 01 exhibited loss of adhesion (peeling). In areas with loss of adhesion, the base metal did not exhibit any degradation that would affect the cask's intended function. The majority of the base metal was clean, however some corrosion and corrosion product stains were observed, mainly in areas where the epoxy coating itself was cracking. In those areas, the base metal did not have observable loss of material (no depth). Additionally, the concrete under the cask exhibited no visual signs of degradation.

With the protective cover removed, inspection of the area underneath the cover of Cask 01 was performed. During this inspection, no subcomponents within the scope of License Renewal exhibited any evidence of degradation. The observable area of the lid and lid bolt heads had no indication of corrosion. A coating of rust was found on the cask flange at the protective cover interface. This rust coating originated on the carbon steel protective cover, was deposited on the cask flange, and was easily removed. The removal of this coating revealed no degradation to the stainless steel overlay surface of the cask flange and no corrosion between the lid and flange in the main lid seal area.

The neutron shield bolts were removed, inspected, and observed to have no indication of corrosion with the N-5000 lubricant still intact on the threads. The neutron shield protective coating exhibited no signs of corrosion.

The protective cover was found to have uniform corrosion on the flange sealing surface outside (external side) of the O-ring seal. There was minor corrosion around the protective cover bolt holes where the bolt heads had broken the epoxy coating due to friction upon installation. The underside of the protective cover dome had no signs of degradation. The protective cover O-ring seal remained in acceptable condition with the exterior coating on the protective cover exhibiting checking on approximately 15% to 20% of the surface area.

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CoC 1027 Renewal Application Revision 0 / April 2020 Inspection of the area underneath the protective cover of Cask 13 was also performed with the protective cover removed. During this inspection, no subcomponents within the scope of License Renewal exhibited any evidence of degradation. The observable area of the lid and lid bolt heads had no indication of corrosion. The stainless steel flange overlay had only small stains where rust from the protective cover was deposited. The stains were removed and there was no indication of corrosion on the observable area of the flange and no corrosion was observed between the lid and flange near the main lid seal area. The neutron shield bolts were removed and inspected with no indication of corrosion; they also had the N-5000 lubricant still intact on the threads. The neutron shield had two rust stains on the protective coating directly below the access cover with one stain approximately six inches in diameter and the other approximately two inches in diameter. The protective cover was found to have corrosion on the interior. The corrosion appears to have started at the interior face of the access cover opening where the stainless steel overpressure system piping welded to the access plate made contact with the protective cover. The protective cover dome had evidence of corrosion in the area where it connected to the access plate. The access plate itself had corrosion on the entire interior surface excluding the area that was covered by the rubber gasket. However, none of these subcomponents exhibiting corrosion are within the scope of license renewal.

During the baseline inspections of Casks 01 and 13, the accessible areas of the casks were also inspected. The upper trunnions of Cask 01 exhibited some corrosion product stains on the top of the trunnion shaft. There was no indication of corrosion on all other areas inspected on Cask 01 and Cask 13.

3C.5.2 North Anna ISFSI Pre-Application Inspection In the fall of 2015, Virginia Electric and Power Company conducted a license renewal pre-application inspection [3C-12] of TN-32 casks at their North Anna Power Station ISFSI. In addition to the accessible exterior surfaces of a cask, the inspection also included visual inspections on the bottom of a cask and beneath the protective cover of another cask. The results of the pre-application inspection are summarized in [3C-12].

Page 3C-6

CoC 1027 Renewal Application Revision 0 / April 2020 The results of the bottom inspection revealed areas where portions of the protective coating exhibited loss of adhesion and areas where the coating adhered to the concrete.

Rust stains were also observed. There was no detectable loss of material from the base metal. Visual inspection of the concrete beneath the cask revealed no detectable degradation, other than shrinkage cracking that occurred during concrete curing. The inside and outside of the protective cover and subassembly were found to be acceptable and no blemishes or rust stains were noted. Surface corrosion stains were noted on four of the protective cover bolt holes, but no detectable loss of material. The upper and lower trunnions exhibited surface corrosion on the top of each trunnion. The bottom half of each trunnion appeared to have a narrow ring of surface corrosion as well. There was no detectable loss of material from the trunnions. The visible portions of the lid, lid bolts, top neutron shield enclosure, neutron shield bolts, and overpressure system components were all in acceptable condition, i.e., no detectable loss of material from the base metal. Four of the twelve protective cover bolts and two of the four neutron shield bolts were inspected using VT-1 NDE techniques. Two protective cover bolts were noted to have surface corrosion stains on the bolt shank, but no detectable loss of material. The stains were easily removed with Scotch-Brite'. No corrosion or rust stains were noted on the neutron shield bolts.

Page 3C-7

CoC 1027 Renewal Application Revision 0 / April 2020 3C.6 Licensee Event Reports A search was performed of the NRCs Licensee Event Report database using the following key words:

  • Cask
  • Dry Fuel Storage
  • TN-32
  • TN-40
  • TN-68 No Licensee Event Reports were found associated with aging-related degradation of the TN-32, TN-40 or TN-68 dry storage cask systems.

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CoC 1027 Renewal Application Revision 0 / April 2020 3C.7 Vendor-Issued Safety Bulletins Table 3C-3 provides a summary of the TN Americas LLC Users Group (TNUG)

Technical Bulletins. As shown in Table 3C-3, most conditions were either event-driven or involve subcomponents that are not in-scope for the renewal of CoC 1027.

The few conditions that were aging-related involved aging mechanisms/effects already identified as requiring management, e.g., galvanic corrosion and chloride-induced stress corrosion cracking (CISCC).

Page 3C-9

CoC 1027 Renewal Application Revision 0 / April 2020 3C.8 NRC Generic Communications 3C.8.1 NRC Information Notices The following NRC information notices (INs) were found to be related to aging of dry fuel storage systems.

NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction

[3C-2]

This IN addresses the occurrence of alkali-silica reaction (ASR)-induced concrete degradation on Seismic Category I structures at a nuclear power plant. ASR is an aging mechanism that is evaluated in the AMR for concrete.

NRC Information Notice 2012-20: Potential Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and Maintenance of Dry Cask Storage System Canisters [3C-3]

Several failures in austenitic stainless steels have been attributed to CISCC. The components that have failed at nuclear power plants because of this failure mechanism are made from the same types of austenitic stainless steels typically used in dry cask storage systems. CISCC is a specialized form of a stress corrosion crack, which is an aging mechanism that is evaluated in the AMR for stainless steel.

NRC Information Notice 2013-07: Premature Degradation of Spent Fuel Storage Cask Structures and Components from Environmental Moisture [3C-4]

This IN describes OE on premature degradation of structures and components important-to-safety (ITS) during spent nuclear fuel storage operations due to environmental moisture. The instances described in this IN illustrate how the intrusion of water can potentially decrease the effective life of both the structures and components of a spent fuel storage system. In one instance, the presence of water not only caused chemical degradation through oxidation of one metal, but it also facilitated the formation of a galvanic cell between two dissimilar metals, which contributed to the degradation of the secondary confinement barrier of the storage system. In another instance, water contributed to an accelerated aging process of the concrete structures of the spent fuel storage system through the freeze-thaw aging mechanism. The AMR assumes the components will be exposed to environmental moisture when in air-outdoor and groundwater/soil environments.

3C.8.2 NRC Bulletins NRC Bulletin 96-04 Chemical, Galvanic or other Reactions in Spent Fuel Storage and Transportation Casks [3C-5]

NRC Bulletin 96-04 addresses the generation of hydrogen due to the reaction internal materials with spent fuel pool water during loading. The specific concern associated with NRC Bulletin is not related to aging of dry fuel storage systems and does not need to be addressed in the AMR.

Page 3C-10

CoC 1027 Renewal Application Revision 0 / April 2020 No other NRC Bulletins were found related to aging of dry fuel storage systems.

Page 3C-11

CoC 1027 Renewal Application Revision 0 / April 2020 3C.9 Updated Consensus Codes, Standards, or Guides 3C.9.1 Aging Management Reviews The AMR conducted for the renewal of CoC 1027 is based on the aging mechanisms/effects described in NUREG-2214 [3C-8]. Therefore, the AMR is based on the version/edition/revision of the codes, standards, and guides referenced in NUREG-2214.

3C.9.2 Aging Management Programs The AMPs for the renewal of CoC 1027 will use the most up-to-date (as of the time the AMPs are developed) version/edition/revision of the applicable codes, standards, and guidelines.

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CoC 1027 Renewal Application Revision 0 / April 2020 3C.10 Applicable Industry Initiatives 3C.10.1 High Burnup Fuel Demonstration Project The Electrical Power Research Institute (EPRI) and Department of Energy (DOE) are conducting a joint High Burnup Dry Storage Cask Research and Development Project (HDRP) [3C-7] that monitors the performance of high burnup (HBU) fuel in dry storage. The HDRP is a program designed to collect data from an SNF storage system containing HBU fuel in a dry helium environment. The program entails loading and storing a TN-32B bolted lid cask (the Research Project Cask) at Dominion Virginia Powers North Anna Power Station with intact HBU fuel (of nominal burnups ranging between 53 GWd/MTU and 58 GWd/MTU). The fuel used in the program includes four kinds of cladding (Zircaloy-4, low-tin Zircaloy-4, ZIRLOTM, and M5TM). The Research Project Cask was loaded and placed in service in November 2017[3C-13].

3C.10.2 Chloride-Induced Stress Corrosion Cracking EPRI has undertaken various initiatives to improve the understanding of the CISCC cracking phenomena in terms of susceptibility [3C-9], risk [3C-10], and aging management guidance [3C-11]. While this initiative is focused on dry shielded canisters (DSC), it was reviewed and determined not to be applicable (due differences in materials, configurations, and lack of stainless steel to stainless steel welds) to the AMP for the TN-68 casks.

Page 3C-13

CoC 1027 Renewal Application Revision 0 / April 2020 3C.11 Conclusion The review of various sources of OE did not identify any aging mechanism and/or effects that were not already identified in NUREG-2214 [3C-8]. In addition, no incidents were identified where aging effects lead to the loss of intended safety functions of SSCs of TN bolted casks. Therefore, it is concluded that the effects of aging will be managed adequately so that the SSCs intended safety functions will be maintained during the PEO.

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CoC 1027 Renewal Application Revision 0 / April 2020 3C.12 References 3C-1 Sandia Report SAND2017-2306, Analysis of Corrosion Residue collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask, March, 2017.

3C-2 U.S. Nuclear Regulatory Commission, Information Notice 2011-20, Concrete Degradation by Alkali-Silica Reaction, Office of Nuclear Material Safety and Safeguards, November 18, 2011.

3C-3 U.S. Nuclear Regulatory Commission, Information Notice 2012-20, Potential Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and Maintenance of Dry Cask Storage System Canister, November 14, 2012.

3C-4 U.S. Nuclear Regulatory Commission, Information Notice 2013-07, Premature Degradation of Spent Fuel Storage Cask Structures and Components from Environmental Moisture, April 16, 2013.

3C-5 U.S. Nuclear Regulatory Commission, Bulletin 96-04, Chemical, Galvanic or Other Reactions in Spent Fuel Storage and Transportation Casks, July 5, 1996.

3C-6 Letter L-PI-15-082, from Scott Sharp (NSPM), to Document Control Desk (NRC),

Supplement to Prairie Island Independent Spent Fuel Storage Installation License Renewal Application - Revised Aging Management Plan (TAC No. L24592), October 12, 2015, (Adams Accession Number ML15285A007).

3C-7 Electric Power Research Institute, High Burnup Dry Storage Cask Research and Development Project: Final Test Plan, Rev. 0, DE-NE-0000593, February 27, 2014.

3C-8 U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Process in Storage (MAPS) Report, July 2019.

3C-9 Electric Power Research Institute, Susceptibility Assessment Criteria for Chloride-Induced Stress Corrosion Cracking (CISCC) of Welded Stainless Steel Canisters for Dry Storage Cask Systems, EPRI Report 3002005371, September 2015.

3C-10 Electric Power Research Institute, Dry Cask Storage Welded Stainless Steel Canister Breach Consequence Analysis Scoping Study, EPRI Report 3002008192, November 2017.

3C-11 Electric Power Research Institute, Aging Management Guidance to Address Potential Chloride-Induced Stress Corrosion Cracking of Welded Stainless Steel Canisters, EPRI Report 3002008193, March 2017.

3C-12 North Anna ISFSI - License, No SNM-2507, Renewal Application, (Adams Accession Number ML16153A140).

3C-13 Federal Register Notice for the Final Environmental Assessment for the North Anna ISFSI License Renewal, January 29, 2018, (Adams Accession Number ML17312A606).

Page 3C-15

CoC 1027 Renewal Application Revision 0 / April 2020 Proprietary Information on Pages 3C-16 through 3C-21 Withheld Pursuant to 10 CFR 2.390 Page 3C-16

CoC 1027 Renewal Application Revision 0 / April 2020 CHAPTER 4 AGING MANAGEMENT PROGRAMS CONTENTS 4.1 Introduction .................................................................................................................... 4-1 4.2 Aging Management Program Elements ....................................................................... 4-2 4.3 TN-68 Aging Management Program............................................................................ 4-4 4.3.1 TN-68 AMP - Scope of Program ........................................................................ 4-4 4.3.2 TN-68 AMP - Preventive Actions ....................................................................... 4-4 4.3.3 TN-68 AMP - Parameters Monitored or Inspected ............................................. 4-4 4.3.4 TN-68 AMP - Detection of Aging Effects .......................................................... 4-5 4.3.4.1 TN-68 AMP - Interseal Pressure Monitoring ....................................... 4-5 4.3.4.2 TN-68 AMP - Radiation Monitoring ................................................... 4-5 4.3.4.3 TN-68 AMP - Visual Inspections......................................................... 4-6 4.3.5 TN-68 AMP - Monitoring and Trending............................................................. 4-6 4.3.6 TN-68 AMP - Acceptance Criteria ..................................................................... 4-7 4.3.6.1 TN-68 AMP - Interseal Pressure .......................................................... 4-7 4.3.6.2 TN-68 AMP - Radiation Monitoring ................................................... 4-7 4.3.6.3 TN-68 AMP - Visual Inspections......................................................... 4-7 4.3.7 TN-68 AMP - Corrective Actions ....................................................................... 4-8 4.3.8 TN-68 AMP - Confirmation Process .................................................................. 4-8 4.3.9 TN-68 AMP - Administrative Controls .............................................................. 4-8 4.3.10 TN-68 AMP - Operating Experience .................................................................. 4-8 4.4 Storage Pad Aging Management Program ................................................................ 4-10 4.4.1 Storage Pad AMP - Scope of Program .............................................................. 4-10 4.4.2 Storage Pad AMP - Preventive Actions ............................................................ 4-10 4.4.3 Storage Pad AMP - Parameters Monitored or Inspected .................................. 4-10 4.4.4 Storage Pad AMP - Detection of Aging Effects ............................................... 4-11 4.4.5 Storage Pad AMP - Monitoring and Trending .................................................. 4-11 4.4.6 Storage Pad AMP - Acceptance Criteria ........................................................... 4-12 4.4.7 Storage Pad AMP - Corrective Actions ............................................................ 4-13 4.4.8 Storage Pad AMP - Confirmation Process ........................................................ 4-13 4.4.9 Storage Pad AMP - Administrative Controls .................................................... 4-13 Page 4-i

CoC 1027 Renewal Application Revision 0 / April 2020 4.4.10 Storage Pad AMP - Operating Experience ........................................................ 4-13 4.5 High Burnup Fuel Aging Management Program ..................................................... 4-15 4.5.1 HBU Fuel AMP - Scope of Program ................................................................ 4-15 4.5.2 HBU Fuel AMP - Preventive Actions ............................................................... 4-15 4.5.3 HBU Fuel AMP - Parameters Monitored or Inspected ..................................... 4-16 4.5.4 HBU Fuel AMP - Detection of Aging Effects .................................................. 4-16 4.5.5 HBU Fuel AMP - Monitoring and Trending ..................................................... 4-16 4.5.6 HBU Fuel AMP - Acceptance Criteria ............................................................. 4-17 4.5.7 HBU Fuel AMP - Corrective Actions ............................................................... 4-17 4.5.8 HBU Fuel AMP - Confirmation Process........................................................... 4-17 4.5.9 HBU Fuel AMP - Administrative Controls ...................................................... 4-18 4.5.10 HBU Fuel AMP - Operating Experience .......................................................... 4-18 4.6 References ..................................................................................................................... 4-19 LIST OF TABLES Table 4-1 Subcomponents Within Scope of TN-68 AMP ................................................. 4-20 Table 4-2 Subcomponents Within Scope of Storage Pad AMP ......................................... 4-23 Table 4-3 HBU Fuel AMP Tollgate ................................................................................... 4-25 Page 4-ii

CoC 1027 Renewal Application Revision 0 / April 2020 4.1 Introduction The purpose of an aging management program (AMP) is to monitor and control the degradation of structures, systems, and components (SSCs) to ensure that no aging effects result in a loss of intended safety function of the in-scope SSCs for the period of extended operations (PEO). An effective AMP prevents, mitigates, or detects aging effects and provides for the prediction of the extent of the effects of aging and timely implementation of corrective actions before there is a loss of intended safety function.

AMPs are based on the results of the aging management review (AMR) for the TN-68 casks, storage pad, and high burnup (HBU) fuel presented in Chapter 3. The tables in Chapter 3 summarize the results of the AMR and identify the aging management activity (AMA) credited for managing each aging effect and aging mechanism for each component or subcomponent evaluated in the AMR.

The AMPs will apply under extended Certificate of Compliance (CoC) No. 1027 terms, that is, implementation of the AMPs will be a CoC condition for the TN-68 dry storage cask system components in service after the initial 20-year period at an independent spent fuel storage installation (ISFSI). The requirement for implementation of the proposed AMPs will be set forth under the terms and conditions of the proposed renewed certificate as described in Attachment B.

The AMPs developed to manage aging effects for the PEO are:

CoC 1027 Renewal Application Revision 0 / April 2020 4.2 Aging Management Program Elements The structure of the AMPs is consistent with the 10 program elements described in NUREG-1927 [4-1], as follows:

1. Scope of the program: The scope of the program includes the specific SSCs and subcomponents subject to the AMP and the intended safety functions to be maintained. In addition, the element states the specific materials, environments, and aging mechanisms and effects to be managed.
2. Preventive actions: Preventive actions used to prevent aging or mitigate the rates of aging for SSCs.
3. Parameters monitored or inspected: This element identifies the specific parameters that will be monitored or inspected and describes how those parameters will be capable of identifying degradation or potential degradation before there is a loss of intended safety function.
4. Detection of aging effects: This element includes inspection and monitoring details, including method or technique (i.e., visual, volumetric, surface inspection), frequency, sample size, data collection, and timing of inspections to ensure timely detection of aging effects. In general, the information in this element describes the when, where, and how of the AMP (i.e., the specific aspects of the activities to collect data as part of the inspection or monitoring activities).

Accessible areas are defined as surfaces of in-scope SSCs and subcomponents that can be visually inspected by direct means without the aid of ladders, scaffolding, removal of protective cover, lifting of cask, or climbing, i.e., visually inspected while inspector is standing on the ground.

Normally non-accessible areas are defined as surfaces (or portions of surfaces) of in-scope SSCs and subcomponents that can be visually inspected directly or by remote means with the aid of ladders, use of scaffolding, removal of protective cover, or lifting of cask.

Inaccessible areas are defined as surfaces of in-scope SSC and subcomponents that cannot be inspected without major excavation or disassembly, e.g., below grade concrete surfaces or embedded reinforcing bars.

5. Monitoring and trending: This element describes how the data collected will be evaluated. This includes an evaluation of the results against the acceptance criteria and an evaluation regarding the rate of degradation to ensure that the timing of the next scheduled inspection will occur before there is a loss of intended safety function.
6. Acceptance criteria: Acceptance criteria, against which the need for corrective action will be evaluated, ensures that the intended safety functions and the approved design bases of the SSC are maintained during the PEO.

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CoC 1027 Renewal Application Revision 0 / April 2020

7. Corrective actions: Corrective actions are the measures taken when the acceptance criteria are not met. Timely corrective actions, including root cause determination and prevention of recurrence for significant conditions adverse to quality, are critical for maintaining the intended safety functions of the SSCs during the PEO.
8. Confirmation process: This element verifies that preventive actions are adequate and that effective appropriate corrective actions have been completed. The confirmation process is commensurate with the general licensee Quality Assurance (QA) Program approved under 10 CFR Part 50, Appendix B. The QA Program ensures that the confirmation process includes provisions to preclude repetition of significant conditions adverse to quality.
9. Administrative controls: Administrative controls provide a formal review and approval process in accordance with an approved QA program.
10. Operating experience (OE): The OE element of the program supports a determination that the effects of aging will be adequately managed so that the SSCs intended safety functions will be maintained during the PEO. Operating experience provides justification for the effectiveness of each AMP program element and critical feedback for enhancement.

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CoC 1027 Renewal Application Revision 0 / April 2020 4.3 TN-68 Aging Management Program 4.3.1 TN-68 AMP - Scope of Program This program visually inspects and monitors the condition of the TN-68 subcomponents listed in Table 4-1. The table also lists the material and environments for each subcomponent along with the aging mechanisms and aging effects to be managed. The following aging effects and mechanism will be managed via this AMP:

  • Steel Loss of material due to general, pitting, crevice, and galvanic corrosion Loss of preload due to stress relaxation of bolts
  • Stainless Steel Loss of material due to pitting, crevice, and galvanic corrosion
  • Polymers Shrinkage/cracking due to thermal aging and radiation embrittlement
  • Aluminum Loss of material due to general, pitting, crevice, and galvanic corrosion 4.3.2 TN-68 AMP - Preventive Actions The program is a condition-monitoring program that does not include preventive actions.

4.3.3 TN-68 AMP - Parameters Monitored or Inspected The TN-68 AMP consists of monitoring the interseal pressure, radiation monitoring, and visual inspections.

Interseal Pressure Monitoring The interseal pressure of the TN-68 dry storage cask seals is monitored to verify the integrity of the TN-68 dry storage cask seals. The interseal region is pressurized via the overpressure system to provide indication of cask seal integrity. A reduction of interseal pressure could indicate leakage due to loss of material of the seals or loss of preload of the bolts and would occur before there is a loss of the confinement intended safety function.

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CoC 1027 Renewal Application Revision 0 / April 2020 Radiation Monitoring Periodic radiation monitoring, both gamma and neutron monitoring, will be conducted to ensure that there is no loss of the shielding intended safety function (i.e., met the requirements of 10 CFR Part 20 and 10 CFR Part 72.104) due to loss of material of the steel and stainless steel subcomponents or due to shrinkage/cracking of the polymer subcomponents. Trending the results of the radiation monitoring will enable detection of aging-related degradation before there is a loss of the shielding intended safety function.

Visual Inspections Periodic visual inspections will be performed on the TN-68 casks looking for loss of material for steel and stainless steel subcomponents, i.e., corrosion. The frequency of these inspections will ensure that the loss of material is detected prior to the loss of an intended safety function.

4.3.4 TN-68 AMP - Detection of Aging Effects This program manages the TN-68 aging effects by monitoring the interseal pressure, radiation monitoring, and visual inspections.

4.3.4.1 TN-68 AMP - Interseal Pressure Monitoring The interseal pressure is monitored by measuring the pressure in the overpressure system. The TN-68 AMP utilizes the same equipment, methods, and frequency used to comply with Technical Specification 3.1.5, Cask Interseal Pressure. Technical Specification Surveillance Requirement SR 3.1.5.1 requires verification that the cask interseal helium pressure is above 3.0 atm absolute every seven days.

4.3.4.2 TN-68 AMP - Radiation Monitoring Detection of gamma and neutron radiation is accomplished by the placement of thermoluminescent dosimeters (TLDs) at the ISFSI perimeter fence, or between the ISFSI and locations used to show compliance with 10 CFR 20.1301 and 10 CFR 72.104. While the TLDs may not be capable of detecting all neutrons (i.e., the very high energy neutrons) they are effective in detecting adverse trends in neutron dose rates.

Thermoluminescent dosimeter readings are obtained quarterly. Results from this monitoring activity provide a means to detect deterioration of the TN-68 dry storage cask gamma and neutron shielding due to loss of material, shrinkage, or cracking before the limits in 10 CFR 20.1301 and 10 CFR 72.104 are exceeded.

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CoC 1027 Renewal Application Revision 0 / April 2020 4.3.4.3 TN-68 AMP - Visual Inspections Accessible surfaces of all TN-68 casks will be visually inspected on an annual basis (plus 25% allowed by Technical Specification SR 3.0.2). Visual (direct or by remote means) inspections of opportunity (e.g., in the event a TN-68 cask is lifted or a protective cover is removed) will be performed on the surfaces of in-scope subcomponents in the normally non-accessible areas. A scheduled visual inspection of normally non-accessible areas of a lead TN-68 cask will be performed within two years prior to 20 years of the first loaded TN-68 cask being placed in storage, or no later than eighteen months after the effective date of the CoC renewal, whichever is later (i.e., a base line inspection), and on a frequency of every 20 +/- 1 years thereafter.

The lead cask is defined as the cask that has been in-service the longest.

Note: Eighteen months allows one year for development of the infrastructure for AMP implementation plus six months to complete subsequent baseline inspections.

The visual inspections are looking for loss of material indicated by corrosion or rust stains of in-scope subcomponents. Since the identification of corrosion and rust stains is a simple skill, non-destructive examination (NDE) qualifications are not required to perform these visual inspections.

Visual inspection of the flange stainless steel weld overlay and carbon steel subcomponents serve as a leading indicator for stainless and carbon steel subcomponents that are not visible during the opportunistic and scheduled inspections, e.g., seals, vent and drain port covers. Similarly, this AMP relies upon these leading indicators to manage aging effects of installed bolts, such that removal of bolts for inspection is not required.

4.3.5 TN-68 AMP - Monitoring and Trending The inspections and monitoring activities in this AMP are performed periodically in order to identify areas of degradation. Conditions that do not meet the acceptance criteria are entered into the licensees corrective action program. Visual inspections appropriately consider cumulative OE from previous inspections and assessments, in order to monitor and trend the progression of aging effects over time. Data taken from these inspections and monitoring activities is to be monitored by comparison to past site data taken as well as comparison to industry OE, including data gathered by the Institute of Nuclear Power Operations (INPO) Aging Management INPO Database (AMID) as discussed in Nuclear Energy Institute (NEI) 14-03 [4-2].

As described in Section 4.3.4.3, one lead TN-68 cask is to be selected for the baseline inspection and subsequent inspections of normally non-accessible areas (e.g., the bottom of the cask and under the protective weather cover). If the lead TN-68 cask is not available for subsequent inspections (e.g., has been shipped off-site), another TN-68 cask is to be selected for a new baseline inspection following the considerations/criteria in Section 4.3.4.3.

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CoC 1027 Renewal Application Revision 0 / April 2020 The quarterly gamma and neutron radiation readings from the TLDs on the ISFSI perimeter fence, or between the ISFSI and locations used to show compliance with 10 CFR 20.1301 and 10 CFR 72.104, are trended to determine if there is an annual increasing trend.

4.3.6 TN-68 AMP - Acceptance Criteria The TN-68 AMP acceptance criteria ensure that the particular structure and component intended functions are maintained under the existing design basis conditions during the PEO. If any of the acceptance criteria below are not met, further evaluation is required through the licensees corrective action program.

4.3.6.1 TN-68 AMP - Interseal Pressure The acceptance criterion for interseal pressure monitoring is the limit specified in Technical Specification 3.1.5, i.e., the cask interseal pressure shall be maintained at a pressure of a least 3.0 atm abs.

4.3.6.2 TN-68 AMP - Radiation Monitoring The acceptance criterion for radiation monitoring is the absence of an annual increasing trend in neutron or gamma quarterly TLD readings at the ISFSI perimeter fence, or between the ISFSI and locations used to show compliance with 10 CFR 20.1301 and 10 CFR 72.104.

4.3.6.3 TN-68 AMP - Visual Inspections To ensure that an evaluation is performed before there is a loss of intended functions due to a loss of material, the acceptance criteria for the visual inspections are.

  • No observed corrosion
  • No rust stains on steel or stainless steel surfaces
  • No rust stains on the concrete pad Page 4-7

CoC 1027 Renewal Application Revision 0 / April 2020 4.3.7 TN-68 AMP - Corrective Actions Site QA procedures, review and approval processes, and administrative controls are implemented according to the requirements of 10 CFR Part 50 Appendix B. The licensees corrective action program ensures that conditions adverse to quality are promptly identified and corrected, including root cause determination and prevention of recurrence. Deficiencies are either corrected or are evaluated to be acceptable for continued service through engineering analysis, which provides reasonable assurance that the intended safety function is maintained consistent with current licensing basis conditions. Extent of condition investigation per the licensees corrective action program may cause additional inspections through means of a different method, increased inspection frequency, and/or expanded inspection sample size. Engineering analysis performed to resolve deficiencies shall be based on the same methods used to provide reasonable assurance that the intended safety function is maintained consistent with the current licensing and safety basis.

4.3.8 TN-68 AMP - Confirmation Process The confirmation process will be commensurate with the general licensee QA program approved under 10 CFR Part 50, Appendix B. The QA program ensures that the confirmation process includes provisions to verify that appropriate corrective actions have been completed and are effective. It also contains provisions to preclude repetition of significant conditions adverse to quality.

4.3.9 TN-68 AMP - Administrative Controls Administrative controls under the CoC holder or licensees QA procedures and corrective action program provide a formal review and approval process.

Administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, and will continue for the PEO. Licensees and the CoC holder use the 10 CFR Part 72 regulatory requirements to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the U.S. Nuclear Regulatory Commission (NRC).

Individual events and conditions not rising to the level of NRC reportability based on the criteria in 10 CFR Part 72 are communicated to the CoC holder as outlined in NEI 14-03 [4-2].

4.3.10 TN-68 AMP - Operating Experience Appendix 3C documents the review of various sources of OE relevant to the TN-68 dry storage cask system. This review included inspections of TN-32 and TN-40 casks that have been in service for several years. While the review identified several conditions that were aging-related, no incidents were identified where aging effects lead to the loss of intended safety functions of a SSC. This OE review supports the conclusion that the effects of aging will be managed adequately so that the SSCs intended safety functions will be maintained during the PEO.

Page 4-8

CoC 1027 Renewal Application Revision 0 / April 2020 This AMP will be updated, as necessary, to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research. Future plant-specific and industry aging management and aging-related OE are captured through the licensees OE review process. The ongoing review of both plant-specific and industry OE will continue through the PEO to ensure that this AMP continues to be effective in managing the identified aging effects.

Page 4-9

CoC 1027 Renewal Application Revision 0 / April 2020 4.4 Storage Pad Aging Management Program 4.4.1 Storage Pad AMP - Scope of Program This program visually inspects the surfaces of the storage pad subcomponents listed in Table 4-2. The table also lists the material and environments for each subcomponent along with the aging mechanisms and aging effects to be managed. The following aging effects/mechanisms will be managed via this AMP:

  • Steel Loss of material due to general, pitting, and crevice corrosion
  • Concrete Loss of material due to freeze-thaw, aggressive chemical attack, corrosion of reinforcing steel, delayed ettringite formation 1, salt scaling, and microbiological degradation Cracking due to freeze-thaw, reaction with aggregates, differential settlement, aggressive chemical attack, corrosion of reinforcing steel, and delayed ettringite formation 1 Loss of strength due to reaction with aggregates, aggressive chemical attack, corrosion of reinforcing steel, leaching of calcium hydroxide, delayed ettringite formation 1, and microbiological degradation Reduction of concrete pH due to aggressive chemical attack, leaching of calcium hydroxide, and microbiological degradation Loss of concrete/steel bond due to corrosion of reinforcing steel Increase in porosity and permeability due to leaching of calcium hydroxide and microbiological degradation 4.4.2 Storage Pad AMP - Preventive Actions The program is a condition-monitoring program that does not include preventive actions.

4.4.3 Storage Pad AMP - Parameters Monitored or Inspected The Storage Pad AMP consists of visual inspections to monitor for material degradation.

The following accessible areas of the storage pad will undergo direct visual inspection for the aging effects listed in Table 4-2:

  • The aboveground exposed surface of the storage pad 1

Delayed ettringite formation may be ruled out as a credible aging mechanism by the general licensee based on an ISFSI-specific evaluation.

Page 4-10

CoC 1027 Renewal Application Revision 0 / April 2020 The normally non-accessible areas of the storage pad include:

  • External surfaces of the storage pad under the TN-68 casks The inaccessible areas of the storage pad include:
  • Below-grade surfaces off the storage pad
  • Components embedded in concrete 4.4.4 Storage Pad AMP - Detection of Aging Effects Direct visual inspections utilizing American Concrete Institute (ACI) report ACI-349.3R [4-3], Section 3.5.1 are to be conducted of the above-grade portions of the concrete storage pad, allowing for detection of aging effects from Table 4-2.

For storage pad concrete, crack maps are developed. Dimensioning is documented in photographic records by inclusion of a tape measure/crack gauge, a comparator, or both.

Potential degradation of the below-grade portion of the concrete pad is assessed by results of groundwater sampling at a minimum of three locations in the area of the ISFSI.

The baseline AMP visual inspection and groundwater sampling is to be conducted within two years prior to 20 years of the first loaded TN-68 being placed in storage, or no later than eighteen months after the effective date of the CoC renewal, whichever is later. Subsequent inspections and groundwater sampling are to be conducted every 5 years +/- 1 year following the baseline inspection.

Note: Eighteen months allows one year for development of the infrastructure for AMP implementation plus six months to complete subsequent baseline inspections.

4.4.5 Storage Pad AMP - Monitoring and Trending The inspections and monitoring activities in this AMP are performed periodically in order to identify areas of degradation. Conditions that do not meet the acceptance criteria are entered into the licensees corrective action program. Other conditions that are noted during the inspection and monitoring activities, such as non-conformances, failures, malfunctions, deficiencies, and deviations are addressed in accordance with the licensees practices and expectations. Visual inspections appropriately consider cumulative OE from previous inspections and assessments in order to monitor and trend the progression of aging effects over time. Data taken from these inspections and sampling is to be monitored by comparison to past site data taken, as well as comparison to industry OE, including data gathered by the AMID as discussed in NEI 14-03 [4-2].

Page 4-11

CoC 1027 Renewal Application Revision 0 / April 2020 For storage pad concrete, crack maps are monitored and trended as a means of identifying progressive growth of defects that may indicate degradation due to specific aging effects, such as rebar corrosion. Crack maps and photographic records are compared with those from previous inspections to identify accelerated degradation of the concrete during the PEO.

4.4.6 Storage Pad AMP - Acceptance Criteria Concrete acceptance criteria from ACI 349.3R [4-3] represent acceptable conditions for observed degradation that has been determined to be inactive. These criteria are termed second-tier for structures possessing a concrete cover in excess of the minimum requirements of ACI 349. Inactive degradation can be determined by the quantitative comparison of current observed conditions with that of prior inspections.

If there is a high potential for progressive degradation or propagation to occur at its present or an accelerated rate, the disposition should consider more frequent evaluations of the specific structure or initiation of repair planning.

The following findings from a visual inspection are considered acceptable:

  • Absence of leaching and chemical attack, including microbiological chemical attack
  • Absence of signs of corrosion in the steel reinforcement
  • Drummy areas that cannot exceed the cover concrete thickness in depth
  • Popouts and voids less than 50 mm (2 in.) in diameter or equivalent surface area
  • Scaling less than 30 mm (1.125 in.) in depth
  • Spalling less than 20 mm (0.75 in.) in depth and 200 mm (8 in.) in any dimension
  • Absence of corrosion staining of undefined source on concrete surfaces
  • Passive cracks less than 1 mm (0.04 in.) in maximum width
  • Passive settlement or deflections within the design basis (serviceability limits)
  • Absence of visible signs of deterioration from alkali-aggregate reaction such as excessive out-of-plane expansion, delayed ettringite formation, or other cement/aggregate reaction The acceptance criteria for the groundwater chemistry-sampling program are:
  • pH 5.5
  • Sulfates 1500 ppm If any of the above acceptance criteria are not met, further evaluation is required through the licensees corrective action program.

Page 4-12

CoC 1027 Renewal Application Revision 0 / April 2020 4.4.7 Storage Pad AMP - Corrective Actions Site QA procedures, review and approval processes, and administrative controls are implemented according to the requirements of 10 CFR Part 50 Appendix B. The licensees corrective action program ensures that conditions adverse to quality are promptly identified and corrected, including root cause determination and prevention of recurrence. Deficiencies are either corrected or are evaluated to be acceptable for continued service through engineering analysis, which provides reasonable assurance that the intended safety function is maintained consistent with current licensing basis conditions. Extent of condition investigation per the licensees corrective action program may cause additional inspections through means of a different method, increased inspection frequency and/or expanded inspection sample size.

4.4.8 Storage Pad AMP - Confirmation Process The confirmation process will be commensurate with the general licensee QA program approved under 10 CFR Part 50, Appendix B. The QA program ensures that the confirmation process includes provisions to verify that appropriate corrective actions have been completed and are effective. It also contains provisions to preclude repetition of significant conditions adverse to quality.

4.4.9 Storage Pad AMP - Administrative Controls Administrative controls under the CoC holder or licensees QA procedures and corrective action program provide a formal review and approval process.

Administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, and will continue for the PEO. Licensees and CoC holder use the 10 CFR Part 72 regulatory requirements to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the NRC. Individual events and conditions not rising to the level of NRC reportability based on the criteria in 10 CFR Part 72 are communicated to the CoC holder as outlined in NEI 14-03 [4-2].

4.4.10 Storage Pad AMP - Operating Experience Appendix 3C documents the review of various sources of OE relevant to the TN-68 dry storage casks. This review included inspections of TN-32 and TN-40 casks that have been in service for several years. While the review identified several conditions that were aging-related, no incidents were identified where aging effects lead to the loss of intended safety functions of a TN-68 cask or storage pad. This OE review supports the conclusion that the effects of aging will be managed adequately so that the SSCs intended safety functions will be maintained during the PEO.

Page 4-13

CoC 1027 Renewal Application Revision 0 / April 2020 This AMP will be updated, as necessary, to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research. Future plant-specific and industry aging management and aging-related OE are captured through the licensees OE review process. The ongoing review of both plant-specific and industry OE will continue through the PEO to ensure that this AMP continues to be effective in managing the identified aging effects.

Page 4-14

CoC 1027 Renewal Application Revision 0 / April 2020 4.5 High Burnup Fuel Aging Management Program 4.5.1 HBU Fuel AMP - Scope of Program Fuel stored in a TN-68 cask is limited to an assembly average burnup of 60 GWd/MTU. The cladding materials for the high burnup (HBU) fuel are zirconium-based, and the fuel is stored in a dry helium environment. The program relies on the joint EPRI and Department of Energy (DOE) HBU Dry Storage Cask Research and Development Project (HDRP) [4-4], conducted in accordance with the guidance in Appendix D of NUREG-1927, Revision 1 [4-1] as a surrogate demonstration program for monitoring the performance of HBU fuel in dry storage. The HDRP is a program designed to collect data from a spent nuclear fuel (SNF) storage system containing HBU fuel in a dry helium environment. The program entails loading and storing an AREVA TN32B bolted lid cask (the Research Project Cask) at Dominion Virginia Powers North Anna Power Station with intact HBU fuel (of nominal burnups ranging between 53 GWd/MTU and 58 GWd/MTU). The fuel to be used in the program includes four kinds of cladding (Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5).

The research project cask is licensed to the temperature limits contained in Interim Staff Guidance (ISG) 11, Revision 3 [4-5], and loaded in such a way that the fuel cladding temperature is as close to the limit as practicable.

The parameters of the surrogate demonstration program are applicable to the designbases HBU fuel, as;

  • maximum allowed burnup of the design-bases HBU fuel (i.e., 60.0 GWd/MTU) is on the order of the nominal burnup of the fuel in the surrogate demonstration program (i.e., 58 GWd/MTU),
  • the cladding temperature of the HBU fuel is limited to the values in ISG-11 and the cladding temperature in the surrogate demonstration program is as close to the ISG-11 limits as practicable.

4.5.2 HBU Fuel AMP - Preventive Actions During the initial loading operations of the cask, the design and CoC 1027 Technical Specifications (TS) require that the fuel be stored in a dry inert environment. TS 3.1.1 demonstrates that the cask cavity is dry by maintaining a cavity absolute pressure at or below 4 mbar for at least 30 minutes with the cask isolated from the vacuum pump.

TS 3.1.2 requires that the cask then be backfilled with helium. These two TS requirements ensure that the HBU fuel is stored in an inert environment, thus preventing cladding degradation due to oxidation mechanisms. The cask is loaded in accordance with the criteria of ISG 11, Revision 3 [4-5].

Page 4-15

CoC 1027 Renewal Application Revision 0 / April 2020 4.5.3 HBU Fuel AMP - Parameters Monitored or Inspected The parameters monitored or inspected are as described in the HDRP [4-4].

While the research project cask is on the storage pad, these parameters include temperature measurements at various locations within the cask. Temperature is the key driver for hydride-induced embrittlement and thermal creep.

It is anticipated that eventually the research project cask will be transported to an offsite fuel examination facility, where the cask will be reopened and the fuel visually examined for changes that occurred during drying and storage.

4.5.4 HBU Fuel AMP - Detection of Aging Effects This AMP relies on the HDRP [4-4] as a surrogate demonstration program for monitoring the performance of HBU fuel in dry storage. The program calls for monitoring cask internal temperatures during the drying process and while the cask is in storage on the ISFSI pad. Temperature is the key driver for hydride-induced embrittlement and thermal creep.

After approximately 10 years of storage at the ISFSI site, it is anticipated that the research project cask will be transported to an off-site fuel examination facility. At the fuel examination facility, the cask will be reopened and the fuel visually examined for changes that occurred during drying and storage. Rods will be extracted from the HBU fuel assemblies and nondestructive and destructive examinations will be performed. It is anticipated that these nondestructive and destructive exams will include cladding profilometry (for creep evaluation), rod internal gas pressure, hydride content and orientation, and cladding mechanical testing (i.e., ductility testing). These examinations will be a direct indication of the susceptibility of high burnup fuel to hydride-induced embrittlement and thermal creep.

4.5.5 HBU Fuel AMP - Monitoring and Trending As information/data from a surrogate demonstration program or from other sources (such as testing or research results and scientific analyses) become available, the licensee will monitor, evaluate, and trend the information via its operating experience program and/or its corrective action program to determine what actions should be taken.

Formal evaluations of the aggregate information from a surrogate demonstration program and other available domestic or international operating experience (including data from monitoring and inspection programs, NRC-generated communications, and other information) will be performed at specific points in time during the PEO, as delineated in Section 4.5.10.

Page 4-16

CoC 1027 Renewal Application Revision 0 / April 2020 4.5.6 HBU Fuel AMP - Acceptance Criteria The following acceptance criteria are to be applied to the data obtained from the HDRP [4-4]. If any of the following fuel performance criteria are not met, the condition will be addressed in accordance with licensees corrective action program:

  • Cladding Temperature: The maximum cladding temperature measured is less than or equal to that predicted by the thermal analysis. A benchmarked thermal model against the demonstration test data may be used for the as-loaded configuration to show that calculated maximum cladding temperature is greater than the demonstrations measured maximum cladding temperature.
  • Cladding Creep: Total creep strain extrapolated to the total approved storage duration based on the best fit to the data, accounting for initial condition uncertainty, shall be less than 1%.
  • Confirmation that hydride reorientation has not compromised the ability to retrieve the spent fuel on a single-assembly basis.

4.5.7 HBU Fuel AMP - Corrective Actions Site QA procedures, review and approval processes, and administrative controls are implemented according to the requirements of 10 CFR Part 50, Appendix B. The licensees corrective action program ensures that conditions adverse to quality are promptly identified and corrected, including root cause determination and prevention of recurrence. Deficiencies are either corrected or are evaluated to be acceptable for continued service through engineering analysis, which provides reasonable assurance that the intended safety function is maintained consistent with current licensing basis conditions. Extent of condition investigation per the licensees corrective action program may cause additional inspections through means of a different method, increased inspection frequency, and/or expanded inspection sample size.

4.5.8 HBU Fuel AMP - Confirmation Process The confirmation process will be commensurate with the general licensee QA program approved under 10 CFR Part 50, Appendix B. The QA program ensures that the confirmation process includes provisions to verify that appropriate corrective actions have been completed and are effective. It also contains provisions to preclude repetition of significant conditions adverse to quality.

Page 4-17

CoC 1027 Renewal Application Revision 0 / April 2020 4.5.9 HBU Fuel AMP - Administrative Controls Administrative controls under the CoC holder or licensees QA procedures and corrective action program provide a formal review and approval process.

Administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, and will continue for the PEO. Licensees and CoC holder use the 10 CFR Part 72 regulatory requirements to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the NRC. Individual events and conditions not rising to the level of NRC reportability based on the criteria in 10 CFR Part 72 are communicated to the CoC holder as outlined in NEI 14-03 [4-2].

4.5.10 HBU Fuel AMP - Operating Experience Short-term testing (i.e., laboratory scale testing up to a few months) and scientific analyses examining the performance of HBU fuel have provided a foundation for the technical basis that storage of HBU fuel in the PEO may be performed safely and in compliance with regulations. However, there has been relatively little OE to date with dry storage of HBU fuel. Therefore, the HDRP is used as a surrogate program to monitor and assess data regarding HBU fuel performance to confirm there is no degradation of HBU fuel that would result in an unanalyzed configuration during the PEO.

This AMP will be updated as necessary to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research. Future plant-specific and industry aging management and age-related OE are captured through the licensees OE review process. The ongoing review of both plant-specific and industry OE will continue through the PEO to ensure that this AMP continues to be effective in managing the identified aging effects.

In addition to the ongoing OE review, this AMP requires periodic written evaluations as described in Table 4-3 of the aggregate impact of aging-related HBU fuel OE, research, monitoring, and inspections on the intended safety functions of the in-scope HBU fuel subcomponents (i.e., tollgates). While licensees and TN Americas LLC assess new information relevant to aging management in accordance with normal corrective action and OE programs, tollgates are an opportunity to seek out other information that may be available and to perform an aggregate assessment. Tollgate assessments are not stopping points. No action, other than performing an assessment and addressing relevant findings in the licensees corrective action program, is required to continue TN-68 dry storage cask system operation. Tollgate assessment reports are not required to be submitted to the NRC, but are available for inspection.

Appendix A of NEI 14-03 [4-2] provides guidance on the performance criteria for the tollgate assessments.

Page 4-18

CoC 1027 Renewal Application Revision 0 / April 2020 4.6 References 4-1 U.S. Nuclear Regulatory Commission, NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Revision 1, June 2016 (ADAMS Accession Number ML16179A148).

4-2 Nuclear Energy Institute, NEI 14-03, Format, Content, and Implementation Guidance for Dry Cask Storage Operations-Based Aging Management, Revision 2, December 2016.

4-3 American Concrete Institute, ACI-349.3R, Evaluation of Existing Nuclear Safety Related Concrete Structures, 2018.

4-4 Electric Power Research Institute, High Burnup Dry Storage Cask Research and Development Project: Final Test Plan, Revision 0, DE-NE-0000593, February 27, 2014.

4-5 NRC Spent Fuel Project Office, Interim Staff Guidance 11, Cladding Considerations for the Transportation and Storage of Spent Fuel, Revision 3, November 17, 2003.

Page 4-19

CoC 1027 Renewal Application Revision 0 / April 2020 Table 4-1 Subcomponents Within Scope of TN-68 AMP (3 Pages)

UFSAR Subcomponent Drawing Intended Safety Parts (Part #s) Function(s) (1) Material Group Environment (2) Credible Aging Mechanism Aging Effect 972-70-2 General Corrosion Loss of Material Gamma Shield SH, TH, SR, RT Steel Air-Outdoor (E)

(1) Pitting and Crevice Corrosion Loss of Material General Corrosion Loss of Material 972-70-2 Lid CO, SH, TH, SR Steel Air-Outdoor(E) Pitting and Crevice Corrosion Loss of Material (2)

Galvanic Corrosion Loss of Material 972-70-2 General Corrosion Loss of Material Bottom SH, TH, SR, RT Steel Air-Outdoor (E)

(4) Pitting and Crevice Corrosion Loss of Material 972-70-2 Upper Trunnion SH, SR, RT Stainless Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material (6) 972-70-2 General Corrosion Loss of Material Lower Trunnion SH, SR, RT Steel Air-Outdoor (E)

(7) Pitting and Crevice Corrosion Loss of Material Radial Neutron 972-70-2 Thermal Aging Shrinkage/Cracking SH, TH Polymer Fully Enclosed Shield (9) Radiation Embrittlement Shrinkage/Cracking 972-70-2 General Corrosion Loss of Material Outer Shell SH, TH, SR Steel Air-Outdoor (E)

(10) Pitting and Crevice Corrosion Loss of Material General Corrosion Loss of Material 972-70-2 Protective Cover SH, TH Steel Air-Outdoor Pitting and Crevice Corrosion Loss of Material (11)

Galvanic Corrosion Loss of Material Top Neutron 972-70-2 Thermal Aging Shrinkage/Cracking SH, TH Polymer Fully Enclosed Shield (12) Radiation Embrittlement Shrinkage/Cracking Page 4-20

CoC 1027 Renewal Application Revision 0 / April 2020 Table 4-1 Subcomponents Within Scope of TN-68 AMP (3 Pages)

UFSAR Subcomponent Drawing Intended Safety Parts (Part #s) Function(s) (1) Material Group Environment (2) Credible Aging Mechanism Aging Effect General Corrosion Loss of Material 972-70-2 Pitting and Crevice Corrosion Loss of Material Lid Bolt CO, SH, TH, SR Steel Air-Outdoor (14) Galvanic Corrosion Loss of Material Stress Relaxation Loss of Preload General Corrosion Loss of Material 972-70-2 Lid Seal CO Aluminum Air-Outdoor Pitting and Crevice Corrosion Loss of Material (16)

Galvanic Corrosion Loss of Material Overpressure Port 972-70-2 CO Stainless Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material Cover (18)

General Corrosion Loss of Material Overpressure Port 972-70-2 CO Aluminum Air-Outdoor Pitting and Crevice Corrosion Loss of Material Cover Seal (19)

Galvanic Corrosion Loss of Material 972-70-2 Drain Port Cover CO, SH, TH, SR Stainless Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material (22) 972-70-2 Vent Port Cover CO, SH, TH, SR Stainless Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material (23)

General Corrosion Loss of Material Vent & Drain Port 972-70-2 CO Aluminum Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material Cover Seal (24)

Galvanic Corrosion Loss of Material General Corrosion Loss of Material Vent & Drain Port Pitting and Crevice Corrosion Loss of Material 972-70-2 Cover Bolts (SOC CO, SR Steel Air-Outdoor (25) Galvanic Corrosion Loss of Material HD Cap)

Stress Relaxation Loss of Preload Page 4-21

CoC 1027 Renewal Application Revision 0 / April 2020 Table 4-1 Subcomponents Within Scope of TN-68 AMP (3 Pages)

UFSAR Subcomponent Drawing Intended Safety Parts (Part #s) Function(s) (1) Material Group Environment (2) Credible Aging Mechanism Aging Effect General Corrosion Loss of Material Overpressure Port Pitting and Crevice Corrosion Loss of Material 972-70-2 Cover Bolts (SOC CO, SR Steel Air-Outdoor (26) Galvanic Corrosion Loss of Material HD Cap)

Stress Relaxation Loss of Preload General Corrosion Loss of Material 972-70-2 Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material Flange CO, SH, TH, SR (35) Galvanic Corrosion Loss of Material Stainless Steel Air-Outdoor (E) Pitting and Crevice Corrosion Loss of Material General Corrosion Loss of Material 972-70-2 Pitting and Crevice Corrosion Loss of Material Trunnion Bolt SR, RT Steel Air-Outdoor (E)

(37) Galvanic Corrosion Loss of Material Stress Relaxation Loss of Preload General Corrosion Loss of Material 972-70-2 Helicoil Inserts CO, SR Steel Air-Outdoor Pitting and Crevice Corrosion Loss of Material (45)

Galvanic Corrosion Loss of Material SOC Head Cap 972-70-2 Pitting and Crevice Corrosion Loss of Material SH, SR Stainless Steel Air-Outdoor Screw (47) Galvanic Corrosion Loss of Material 972-70-2 General Corrosion Loss of Material Shield Ring SH Steel Air-Outdoor (48) Pitting and Crevice Corrosion Loss of Material

1. The intended safety functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR), Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT).
2. If the subcomponent has an internal and external surface exposed to different environments, (I) refers to an internal (or towards the interior of the TN-68) environment and (E) refers to an external (or towards the exterior of the TN-68) environment.

Page 4-22

CoC 1027 Renewal Application Revision 0 / April 2020 Table 4-2 Subcomponents Within Scope of Storage Pad AMP (2 Pages)

Intended Subcomponent Safety Material Credible Aging Parts Function(s) (2) Group Environment Mechanism Aging Effect General Corrosion Loss of Material Pitting and Loss of Material Embedded in Crevice Corrosion Steel Concrete Microbiologically Influenced Loss of Material Corrosion Cracking Freeze-Thaw Loss of Material Reaction with Cracking Aggregates Loss of Strength Cracking Aggressive Loss of Strength Chemical Attack Loss of Material Reduction of Storage Pad TH Concrete pH Loss of Concrete/Steel Bond Concrete Air-Outdoor Corrosion of Loss of Material Reinforcing Steel Cracking Loss of Strength Loss of Strength Leaching of Increase in Porosity Calcium and Permeability Hydroxide Reduction of Concrete pH Loss of Material Delayed Ettringite Loss of Strength Formation (1)

Cracking Salt Scaling Loss of Material Page 4-23

CoC 1027 Renewal Application Revision 0 / April 2020 Table 4-2 Subcomponents Within Scope of Storage Pad AMP (2 Pages)

Intended Subcomponent Safety Material Credible Aging Parts Function(s) (2) Group Environment Mechanism Aging Effect Cracking Freeze-Thaw Loss of Material Reaction with Cracking Aggregates Loss of Strength Differential Cracking Settlement Cracking Loss of Strength Aggressive Chemical Attack Loss of Material Reduction of Concrete pH Loss of Concrete/Steel Bond Corrosion of Loss of Material Reinforcing Steel Groundwater/ Cracking Storage Pad TH Concrete Soil Loss of Strength Loss of Strength Leaching of Increase in Porosity Calcium and Permeability Hydroxide Reduction of Concrete pH Loss of Strength Loss of Material Microbiological Increase in Porosity Degradation and Permeability Reduction of Concrete pH Loss of Material Delayed Ettringite Loss of Strength Formation (1)

Cracking Salt Scaling Loss of Material (1)

Delayed Ettringite Formation may be ruled out as a credible aging mechanism by the general licensee based on an ISFSI-specific evaluation.

(2)

The Intended Safety Functions are: Confinement (CO), Radiation Shielding (SH), Sub-Criticality Control (CR),

Structural Integrity (SR), Heat Removal Capability (TH), Retrievability (RT).

Page 4-24

CoC 1027 Renewal Application Revision 0 / April 2020 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 Page 4-25

CoC 1027 Renewal Application Revision 0 / April 2020 ATTACHMENT A CHANGES TO THE COC 1027 UPDATED FINAL SAFETY ANALYSIS REPORT CONTENTS A.1 Introduction ................................................................................................................... A-1 LIST OF TABLES Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal .......................... A-2 Table A-2 TN-68 Aging Management Program .................................................................. A-8 Table A-3 Storage Pad Aging Management Program ........................................................ A-10 Table A-4 High Burnup Fuel Aging Management Program .............................................. A-12 Table A-5 HBU Fuel AMP Tollgate .................................................................................. A-14 Page A-i

CoC 1027 Renewal Application Revision 0 / April 2020 A.1 Introduction The proposed changes to the Certificate of Compliance (CoC) No. 1027 Updated Final Safety Analysis Report (UFSAR) for the TN-68 Dry Storage Cask System in support of the CoC 1027 renewal are discussed and described in this attachment.

Table A-1 provides a list of changed UFSAR pages, a description of each change, and the basis for the change. UFSAR Revision 9 is referenced in the table for the location of the changed UFSAR text. Table A-2, Table A-3, Table A-4, and Table A-5 provide the format and referenced content for new UFSAR Chapter 15 tables.

Page A-1

CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

1 1.1-1 Following the third paragraph, add the following new note: CoC renewal discussion NOTE: CoC 1027 was originally licensed for 20 years. On [mm/dd/yy], the provided in this historical NRC approved renewal of CoC 1027 for an additional 40 years. The aging introductory UFSAR management activities (AMA) associated with this renewal apply to the previously section, with pointers to approved amendments, and future amendments will include an aging AMR results.

management review (AMR) and any resultant, required aging management activities. The current aging management results are detailed in Chapter 15.

1.2 1.2-5 Add heading 1.2.4 Aging Management Program Requirements with only Aging Add CoC renewal AMP Management Program Requirements underlined. requirements, with pointers Add text: Aging management program (AMP) requirements for use of the TN-68 to location of AMP storage system during the period of extended storage operations are contained in requirements.

Section 15.3.

2.3.1 2.3-1 Change fist sentence to read: The TN-68 dry storage cask is designed to provide storage Bounds Extended storage of spent fuel for at least 40 60 years. period.

2.3.2.1 2.3-3 Change the second sentence of the third paragraph to read: The neutron flux is 2.34xl05 Bounds Extended storage n/cm2s (Chapter 14) equivalent to less than 1.54.5x1014 n/cm2 after 2060 years. period.

2.3.2.1 2.3-4 Change the second paragraph to read: The calculations provided in Chapter 7 define the Clarifies that the analysis in monitoring system leakage test rate which ensures that no cavity gas can be released to Chapter 7 was performed the environment nor air admitted to the casks for the first 20 40 year storage period. If for a 20-year period and that needed, the monitoring system may be re-pressurized to maintain design the monitoring system is conditions. All seals are considered collectively in the analysis as the monitoring capable of being re-system pressure boundary. This analysis is performed in accordance with ANSI pressurized.

N14.5(12).

2.4 2.4-1 Change the fifth paragraph to read: Cask activation analyses have been performed to Clarify that the activation quantify specific activity levels of cask materials after years of storage. This analysis analysis may need to be may need to be updated at the time of decommissioning to reflect actual storage updated to reflect more conditions, e.g., years of storage. The following assumptions were made: realistic conditions.

Page A-2

CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

2.5 2.5-1 Change the first sentence of the second paragraph to read: Add the following sentences The temperature limits in to the end of the second paragraph: The maximum total heat generation rate of the ISG-11 ensure that the fuel stored fuel is limited to 30 kW in order to keep the maximum fuel cladding temperature cladding will continue to below the limit necessary to ensure cladding integrity(13) for 40 years storage(13). perform its intended function through the PEO.

Table 2.5-1 - Change the minimum design life to read: 4060 years Bounds extended storage period.

3.3.4 3.3-1 Change the first sentence to read: Materials must maintain the ability to perform their Materials must maintain safety-related functions over at least the cask's 20 year lifetime under the cask's their safety-related functions thermal, radiological, corrosion, and stress environment. regardless of the duration of the casks lifetime.

3.3.4 3.3-1 The effect of fast neutron irradiation of metals is a function of the integrated fast Bounds extended storage neutron flux, which is on the order of 1014 n/cm2 inside the TN-68 after 40 years. period.

7.1.5 7.1-6 Change the fourth paragraph to read: The overpressure system pressure is also Clarifies that the analysis corrected for the corresponding drop in temperature over the first year. At the end of was performed for a 20-year the first year, the overpressure system pressure is 6.34 atm abs (78.5 psig). These period and that the calculations are repeated every year for the 20 years life of the cask. If needed, the monitoring system is monitoring system may be re-pressurized to maintain design conditions. Figure capable of being re-7.1-1 illustrates the pressure drop from the overpressure system to the atmosphere. pressurized.

Figure 7.1-1 also illustrates the pressure drop in the cask cavity due to fuel cooling.

7.2.1 7.2-1 Change the first sentence to read: The TN-68 dry storage cask is designed to provide Bounds extended storage storage of spent* fuel for at least 4060 years. period.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

7.2.2 7.2-1 Change the first paragraph and equation to read: The TN-68 cask cavity's equilibrium Bounds extended storage pressure during normal storage conditions with no fuel rod rupture is 2.2 atm abs (17.6 period.

psig). The internal pressure is determined on the basis that a minimum of 1 atm pressure must exist on the coldest day at the end of life. Pressure variations due to daily and seasonal changes in ambient temperature conditions will be small due to the large thermal capacity of the cask. The initial pressure of 2.2 atm abs assures that at the end of 40 years, on the coldest day (20°F ambient), the internal pressure of the cask is:

Pcavity = 2.20 atm abs x (596°R / 862°R) = 1. 5 atm abs (7.7 psig) 596° R is the average gas cavity temperature after 40 years of storage assuming an external ambient temperature of -20° F and 21.2 kW initial heat load.

Initially, the cavity is pressurized with helium such that the cavity pressure remains above 1 atm abs pressure on the coldest day at the end of life. The Technical Specification minimum allowed TN-68 cask cavity pressure at the end of the helium backfill is 1.8 atm abs. A steady state run of the full cask model described in Section 4.5.1 determines the average cavity gas temperature after completion of the helium backfilling. An ambient temperature of 70°F is considered for this run. The average gas cavity temperature is 350°F (810°R), and is retrieved from the model using the methodology described in Section 4.7.1. The average gas cavity temperature after 60 years of storage assuming -20°F and a heat load of 8.42 kW was determined to be 47.5°F (507°R). After 60 years of storage, the internal pressure of the cask is:

P60 yrs = 1.8 atm abs X (507°R/810°R) = 1.127 atm abs Page A-4

CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

11.1.2.3 11.1-4 Change the third paragraph to read: In this latter case, leakage out of the interspace to Clarifies that the analysis the atmosphere and the cask cavity would occur. This would not result in release of was performed for a 20-year radioactive material from the cask cavity until the pressure fell to the cask cavity period and that the pressure. At the test leak rate of 1 x 10-5 ref cm3/sec, this would not occur during the a monitoring system is 20 year storage period. If needed, the monitoring system may be re-pressurized to capable of being re-maintain design conditions. pressurized.

14.1 14.1-1 Change the sixth paragraph to read: Cask activation analyses have been performed to Clarify that the activation quantify specific activities of cask materials after years of storage. This analysis may analysis may need to be need to be updated at the time of decommissioning to reflect actual storage updated to reflect more conditions, e.g., years of storage. The following assumptions were made: realistic conditions.

15 new Add centered heading 15. AGING MANAGEMENT New UFSAR chapter.

(new) 15.1 new Add heading 15.1 Aging Management Review with only Aging Management New section describing (new) Review underlined. AMR.

Add text: The aging management review (AMR) of the TN-68 dry storage cask system contained in the application for initial Certificate of Compliance (CoC) renewal [15.1] provides an assessment of aging effects that could adversely affect the ability of in-scope structures, systems, and components (SSCs) to perform their intended functions during the period of extended operation. Aging effects, and the mechanisms that cause them, were evaluated for the combinations of materials and environments identified for the subcomponent of the in-scope SSCs based on a review of the Managing Aging Processes in Storage (MAPS) Report

[15.2]. Aging effects that could adversely affect the ability of the in-scope SSCs to perform their safety function(s) require an aging management activity (AMA) to address potential degradation that may occur during the extended storage period.

The AMA may consist of a time-limited-aging-analysis (TLAA) or an aging management program (AMP). TLAAs and AMPs that are credited with managing aging effects during the extended storage period are discussed in Sections 15.2 and 15.3, respectively.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

15.2 new Add heading 15.2 Time-Limited Aging Analyses with only Time-Limited Aging New section discusses (new) Analyses underlined. TLAAs that are credited Add text: A comprehensive review to identify the TLAAs for the in-scope SSCs of with managing aging effects the TN-68 Dry Storage System was performed to determine the analyses that during the extended storage could be credited with managing aging effects over the period of extended period.

operation. The TLAAs identified involved the in-scope SSCs, considered the effects of aging, involved explicit time-limited assumptions, provided conclusions regarding the capability of an SSC to perform its intended function through the operating term, and were contained or incorporated in the design basis. The identified TLAAs were dispositioned by demonstrating the existing analysis remains valid for the PEO or the analysis was updated. The identified TLAA was:

Ensuring cavity pressure remains above one atmosphere on the coldest day at the end of the storage period.

15.3 new Add heading 15.3 Aging Management Program with only Aging Management New section summarizing (new) Program underlined. the aging management Add text: Aging effects that could result in the loss of in-scope SSCs intended program credited with function(s) are managed during the period of extended operation. Many aging managing aging during the effects are adequately managed for the extended storage period using TLAA, as extended storage period.

discussed in Section 15.2. An AMP is used to manage those aging effects that are New tables below provide not managed by TLAA. The AMPs that manage each of the identified aging effects the summary discussion and for all in-scope SSCs include the following: requirements of each TN-68 Aging Management Program individual aging Storage Pad Aging Management Program management program.

High Burnup Fuel Aging Management Program The AMPs are summarized in Tables 15-1, 15-2, and 15-3 . Additional details are available in the CoC 1027 renewal application [15.1]

15.3 new Add table heading Table 15-1 TN-68 Aging Management Program with only New table to summarize this (new) TN-68 Aging Management Program underlined. AMP.

Add the information from Table A-2 Page A-6

CoC 1027 Renewal Application Revision 0 / April 2020 Table A-1 List of UFSAR Changes Associated with CoC 1027 Renewal (6 Pages)

UFSAR UFSAR Description of Change Revision 9 Revision 9 (Newly inserted text is shown as bold and underlined; deleted text is shown by a Basis for Change Section # Page single strike-through.)

15.3 new Add table heading Table 15-2 Storage Pad Aging Management Program with New table to summarize this (new) only Storage Pad Aging Management Program underlined. AMP.

Add the information from Table A-3.

15.3 new Add table heading Table 15-3 High Burnup Fuel Aging Management Program New table to summarize this (new) with only High Burnup Fuel Aging Management Program underlined. AMP.

Add the information from Table A-4.

15.3 new Add table heading Table 15-4 High Burnup Fuel AMP Tollgate with only High New table to summarize this (new) Burnup Fuel AMP Tollgate underlined. HBU Fuel AMP Tollgate.

Add the information from Table A-5.

15.4 new Add heading 15.4 Supplemental Information with only Supplemental New Section for Chapter 15 (new) Information underlined. References.

15.4.1 New Add heading 15.4.1 References with only References underlined. New Section for Chapter 15 (new) References.

15.4.1 New Add the following references: New Section for Chapter 15 (new) [1] (This renewal application). References.

[2] U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Process in Storage (MAPS) Report July 2019.

[3] Electric Power Research Institute, High Burnup Dry Storage Cask Research and Development Project: Final Test Plan, Revision 0, DE-NE-0000593, February 27, 2014.

[4] NRC Spent Fuel Project Office, Interim Staff Guidance 11, Cladding Considerations for the Transportation and Storage of Spent Fuel, Revision 3, November 17, 2003.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-2 TN-68 Aging Management Program (2 Pages)

AMP Element AMP Activity Scope of Program The program visually inspects the external surfaces and monitors the condition of the TN-68 cask subcomponents.

Preventative Actions The program is a condition-monitoring program that does not include preventive actions.

Parameters Monitored or The interseal pressure of the TN-68 dry storage cask seals is monitored to verify the Inspected integrity of the TN-68 dry storage cask seals.

Periodic radiation monitoring, both gamma and neutron monitoring, will be conducted to ensure that there is no loss of the shielding intended safety function.

Periodic visual inspections will be performed on the TN-68 casks looking for loss of material.

Detection of Aging Effects Selection of TN-68 for The interseal pressure monitoring and visual inspections of accessible surfaces are to Inspection/monitoring be performed on all TN-68 casks in service. The visual inspection of the normally non-accessible areas of TN-68 cask is to be conducted on the cask that has been in service the longest.

Inspection Methods: Interseal Pressure Monitoring The interseal pressure is monitoring by measuring the pressure in the overpressure system.

Radiation Monitoring Detection of gamma and neutron radiation is accomplished by placement of thermoluminescent dosimeters (TLDs) at the independent spent fuel storage installation (ISFSI) perimeter fence, or between the ISFSI and locations used to show compliance with 10 CFR 20.1301 and 10 CFR 72.104.

TN-68 Visual Inspection The direct or remote visual inspections are looking for loss of material indicated by corrosion or rust stains of in-scope subcomponents.

Inspection/Monitoring Interseal Pressure Monitoring Timing and Frequency: The interseal pressure is verified every 7 days Radiation Monitoring TLD readings are obtained quarterly TN-68 Visual Inspection Accessible surfaces are inspected annually. Normally non-accessible areas are to be inspected within two years prior to 20 years of the first loaded TN-68 cask being placed in storage, or no later than eighteen months after the effective date of the CoC renewal, whichever is later, on a frequency of every 20 +/- 1 years thereafter, and if an opportunity presents itself.

Monitoring and Data taken from these inspections and monitoring activities is to be monitored by Trending comparison to past site data taken as well as comparison to industry operating experience (OE), including data gathered by the Aging Management INPO Database (AMID).

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-2 TN-68 Aging Management Program (2 Pages)

AMP Element AMP Activity Acceptance Criteria Interseal Pressure The acceptance criterion for interseal pressure monitoring is the limit specified in Technical Specification 3.1.5.

Radiation Monitoring The acceptance criterion for radiation monitoring is the absence of an annual increasing trend in neutron or gamma quarterly TLD readings at the ISFSI perimeter fence, or between the ISFSI and locations used to show compliance with 10 CFR 20.1301 and 10 CFR 72.104.

Visual Inspections The acceptance criteria for the visual inspections are no observed corrosion, no rust stains, and no rust stains on the concrete pad.

Corrective Actions Per the licensees corrective action program Confirmation Process Per the licensees quality assurance program Administrative Controls Per the licensees quality assurance program. The 10 CFR Part 72 regulatory requirements are used to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the NRC.

Operating Experience This AMP will be updated, as necessary, to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-3 Storage Pad Aging Management Program (2 Pages)

AMP Element AMP Activity Scope of Program The program visually inspects the surfaces of the storage pad.

Preventative Actions The program is a condition-monitoring program that does not include preventive actions.

Parameters Monitored or The aboveground, exposed surface of the storage pad will undergo direct visual Inspected inspection for aging effects:

Detection of Aging Effects Areas for Inspection: Above grade portions of the concrete storage pad.

Inspection Methods: Visual inspections of above grade portions of the storage pad. Groundwater sampling at a minimum of three locations in the area of the independent spent fuel storage installation (ISFSI).

Inspection Timing and The baseline AMP visual inspection and groundwater sampling is to be conducted Frequency: within two years prior to 20 years of the first loaded TN-68 being placed in storage, or no later than eighteen months after the effective date of the CoC renewal, whichever is later. Subsequent inspections are to be conducted every 5 +/- 1 years following the baseline inspection.

Monitoring and Trending Data taken from these inspections and sampling is to be monitored by comparison to past site data taken as well as comparison to industry OE, including data gathered by the AMID.

Acceptance Criteria For concrete, the following findings from a visual inspection are considered acceptable without requiring any further evaluation:

Absence of leaching and chemical attack, including microbiological chemical attack Absence of signs of corrosion in the steel reinforcement Drummy areas that cannot exceed the cover concrete thickness in depth Popouts and voids less than 50 mm (2 in.) in diameter or equivalent surface area Scaling less than 30 mm (1-1/8 in.) in depth Spalling less than 20 mm (3/4 in.) in depth and 200 mm (8 in.) in any dimension Absence of corrosion staining of undefined source on concrete surfaces Passive cracks less than 1 mm (0.04 in.) in maximum width Passive settlement or deflections within the design basis (serviceability limits)

Absence of visible signs of deterioration from alkali-aggregate reaction such as excessive out-or-plane expansion, delayed ettringite formation, or other cement/aggregate reaction For the groundwater samples:

pH 5.5 Chlorides 500 ppm Sulfates 1500 ppm Corrective Actions Per the licensees corrective action program Confirmation Process Per the licensees quality assurance program Page A-10

CoC 1027 Renewal Application Revision 0 / April 2020 Table A-3 Storage Pad Aging Management Program (2 Pages)

AMP Element AMP Activity Administrative Controls Per the licensees quality assurance program. The 10 CFR Part 72 regulatory requirements are used to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the NRC.

Operating Experience This AMP will be updated, as necessary, to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-4 High Burnup Fuel Aging Management Program (2 Pages)

AMP Element AMP Activity Scope of Program The program relies on the joint Electric Power Research Institute (EPRI) and Department of Energy (DOE) High Burnup (HBU) Dry Storage Cask Research and Development Project (HDRP) [15.3] as a surrogate demonstration program that monitors the performance of HBU fuel in dry storage.

Preventative Actions During the initial loading operations of the cask, the Technical Specifications (TS) require that the fuel be stored in a dry inert environment. These TS requirements ensure that the HBU fuel is stored in an inert environment, thus preventing cladding degradation due to oxidation mechanisms. The cask is also loaded in accordance with the criteria of ISG 11, Revision 3 [15.4].

Parameters Monitored or While the Research Project Cask is on the storage pad, parameters Inspected monitored include temperature measurements at various locations within the cask.

After several years in storage, it is anticipated that fuel rods will be extracted from the Research Project Cask and exams, including cladding profilometry (for creep evaluation), rod internal gas pressure, hydride content and orientation, and cladding mechanical testing (e.g., ductility testing), will be performed.

Detection of Aging Effects This AMP relies on the HDRP [15.3] as a surrogate demonstration program for monitoring the performance of HBU fuel in dry storage. After approximately 10 years of storage at the ISFSI site, it is anticipated that the Research Project Cask will be transported to an off-site fuel examination facility. At the fuel examination facility, the cask will be reopened and the fuel visually examined for changes that occurred during drying and storage. Rods will be extracted from the HBU fuel assemblies and nondestructive and destructive examinations will be performed. It is anticipated that these nondestructive and destructive exams will include cladding profilometry (for creep evaluation), rod internal gas pressure, hydride content and orientation, and cladding mechanical testing (i.e., ductility testing). These examinations will be a direct indication of the susceptibility of HBU fuel to hydride-induced embrittlement and thermal creep.

Monitoring and Trending As information/data from a surrogate demonstration program or from other sources (such as testing or research results and scientific analyses) become available, the licensee will monitor, evaluate, and trend the information via its operating experience program and/or its corrective action program to determine what actions should be taken.

Acceptance Criteria The following acceptance criteria are to be applied to the data obtained from the HDRP [15.3]:

Cladding Temperature - The maximum cladding temperature measured is less than or equal to that predicted by the thermal analysis.

Cladding Creep - total creep strain extrapolated to the total approved storage duration based on the best fit to the data, accounting for initial condition uncertainty, shall be less than 1%.

Confirmation that hydride reorientation has not compromised the ability to retrieve the spent fuel on a single-assembly basis.

Corrective Actions Per the licensees corrective action program.

Confirmation Process Per the licensees quality assurance program.

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CoC 1027 Renewal Application Revision 0 / April 2020 Table A-4 High Burnup Fuel Aging Management Program (2 Pages)

AMP Element AMP Activity Administrative Controls Per the CoC holder or licensees quality assurance program. The 10 CFR Part 72 regulatory requirements are used to determine if a particular aging-related degradation condition or event identified via OE, research, monitoring, or inspection is reportable to the NRC.

Operating Experience This AMP will be updated as necessary to incorporate new information on degradation due to aging effects identified from plant-specific inspection findings, related industry OE, and related industry research.

In addition to the ongoing OE review, this AMP requires periodic written evaluations, as described in Table 15-4, of the aggregate impact of aging-related HBU fuel OE, research, monitoring, and inspections on the intended safety functions of the in-scope HBU fuel subcomponents (i.e., tollgates).

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CoC 1027 Renewal Application Revision 0 / April 2020 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 Page A-14

CoC 1027 Renewal Application Revision 0 / April 2020 ATTACHMENT B CHANGES TO COC 1027 AND TECHNICAL SPECIFICATIONS CONTENTS B.1 Details of Proposed CoC Changes ................................................................................B-1 B.2 Details of Proposed Technical Specifications Changes...............................................B-3 Page B-i

CoC 1027 Renewal Application Revision 0 / April 2020 B.1 Details of Proposed CoC Changes The following changes are proposed to the Certificate of Compliance (CoC) No. 1027 initial Amendment 0 and Amendment 1 to support the CoC 1027 renewal:

Amendments 0 and 1 Add the following conditions to the CoC:

12. UFSAR UPDATE FOR RENEWED COC The CoC holder shall submit an updated final safety analysis report (UFSAR) to the Commission, in accordance with 10 CFR 72.4, within 90 days of the effective date of the renewal. The UFSAR shall reflect the changes and CoC holder commitments resulting from the review and approval of the renewal of the CoC. The CoC holder shall continue to update the UFSAR pursuant to the requirements of 10 CFR 72.248.
13. 72.212 EVALUATIONS FOR RENEWED COC USE Any general licensee that initiates spent fuel dry storage operations with the TN-68 dry storage cask system after the effective date of the CoC renewal and any general licensee operating a TN-68 dry storage cask system as of the effective date of the CoC renewal, including those that put additional storage systems into service after that date, shall:
a. as part of the evaluations required by 10 CFR 72.212(b)(5), include evaluations related to the terms, conditions, and specifications of this CoC amendment as modified (i.e., changed or added) as a result of the renewal of the CoC;
b. as part of the document review required by 10 CFR 72.212(b)(6),

include a review of the UFSAR changes resulting from the renewal of the CoC and the NRC Safety Evaluation Report related to the renewal of the CoC; and

c. ensure that the evaluations required by 10 CFR 72.212(b)(7) and (8) capture the evaluations and review described in (a.) and (b.) of this CoC condition.

The general licensee shall complete Condition 13 prior to entering the period of extended operation or no later than one year after the effective date of the CoC renewal, whichever is later.

Page B-1

CoC 1027 Renewal Application Revision 0 / April 2020

14. AMENDMENTS AND REVISIONS FOR RENEWED COC All future amendments and revisions to this CoC shall include evaluations of the impacts to aging management activities (i.e., time-limited aging analyses and aging management programs) to ensure that they remain adequate for any changes to SSCs within the scope of renewal.

Page B-2

CoC 1027 Renewal Application Revision 0 / April 2020 B.2 Details of Proposed Technical Specifications Changes The following changes are proposed to the Technical Specifications (TS) associated with CoC 1027 initial Amendment 0 and Amendment 1 to support the CoC 1027 renewal:

Amendment 0 and 1 Revise TS Section 5.2, Programs, as follows:

  • Add the following new section after Section 5.2.3:

5.2.4 Aging Management Program Each general licensee shall have a program to establish, implement, and maintain written procedures for each aging management program (AMP) described in the updated final safety analysis report (UFSAR). The program shall include provisions for changing AMP elements, as necessary, and within the limitations of the approved licensing bases to address new information on aging effects based on inspection findings and/or industry operating experience provided to the general licensee during the renewal period. The program document shall contain a reference to the specific aspect of the AMP element implemented by that program document, and that reference shall be maintained even if the program document is modified.

The general licensee shall establish and implement this program document prior to entering the period of extended operation or no later than one year after the effective date of the CoC renewal, whichever is later. The general licensee shall maintain the program document for as long as the general licensee continues to operate a TN-68 dry storage cask system in service for longer than 20 years.

The general licensee shall obtain CoC holder review and approval of any changes to the Aging Management Program prior to implementation in order to verify that underlying methodologies, calculations or analyses used in determining that no aging effects result in a loss of intended safety functions remain valid.

Page B-3

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