ML23255A254

From kanterella
Jump to navigation Jump to search
Final Safety Evaluation Report for NAC International Multipurpose Canister (NAC-MPC) Storage System Certificate of Compliance No. 1025
ML23255A254
Person / Time
Site: 07201025
Issue date: 09/12/2023
From:
Storage and Transportation Licensing Branch
To:
NAC International
Shared Package
ML23255A244 List:
References
Download: ML23255A254 (1)
v • d • e


Text

FINAL SAFETY EVALUATION REPORT FOR THE NAC INTERNATIONAL MULTI-PURPOSE CANISTER (NAC-MPC) STORAGE SYSTEM CERTIFICATE OF COMPLIANCE NO. 1025 RENEWAL DOCKET NO. 72-1025 Office of Nuclear Material Safety and Safeguards United States Nuclear Regulatory Commission September 12, 2023

CONTENTS INTRODUCTION ...................................................................................................................... VI 1 GENERAL INFORMATION ...............................................................................................1-1 1.1 Certificate of Compliance and Certificate of Compliance Holder Information ............1-1 1.2 Safety Review ..........................................................................................................1-1 1.3 Application Content ..................................................................................................1-2 1.4 Evaluation Findings..................................................................................................1-2 2 SCOPING EVALUATION ..................................................................................................2-1 2.1 Scoping and Screening Methodology .......................................................................2-1 2.1.1 Scoping Process ..........................................................................................2-1 2.1.2 Scoping Results ...........................................................................................2-3 2.1.3 Structures, Systems, and Components Within the Scope of Renewal Review .........................................................................................................2-5 2.2 Evaluation Findings................................................................................................2-11 3 AGING MANAGEMENT REVIEW .....................................................................................3-1 3.1 Review Objective .....................................................................................................3-1 3.2 Aging Management Review Process .......................................................................3-1 3.3 Aging Management Review Results: Materials, Service Environment, Aging Effects, and Aging Management Activities ...............................................................3-1 3.3.1 Supplemental Analyses ..............................................................................3-20 3.3.2 Evaluation Findings ....................................................................................3-21 3.4 Time-Limited Aging Analyses.................................................................................3-22 3.4.1 Fatigue Evaluation of NAC-MPC System Components for Extended Storage ......................................................................................................3-22 3.4.2 Corrosion Analysis of NAC-MPC Steel Components for Extended Storage Operation ......................................................................................3-23 3.4.3 Aging Analysis for NAC-MPC Neutron Absorber and Neutron Shield Components ...............................................................................................3-25 3.4.4 Evaluation of Stainless-Steel Clad Fuel for Fatigue in Storage ...................3-28 3.4.5 Evaluation Findings ....................................................................................3-29 3.5 Aging Management Programs ...............................................................................3-29 3.5.1 Aging Management Tollgates .....................................................................3-36 3.5.2 Evaluation Findings ....................................................................................3-37 4 CHANGES TO CERTIFICATE OF COMPLIANCE AND TECHNICAL SPECIFICATIONS 4-1 5 CONCLUSION ..................................................................................................................5-1 6 REFERENCES .................................................................................................................6-1 ii

TABLES Table Page 2.1-1 SSCs Within and Not Within the Scope of Renewal Review .........................................2-3 2.1-2 Subcomponents Within the Scope of Renewal ReviewTSC ......................................2-6 2.1-3 Subcomponents Within the Scope of Renewal ReviewVCC ......................................2-7 2.1-4 Subcomponents Within the Scope of Renewal ReviewTFR and Transfer Adapter Plate ........................................................................................................................ .2-8 2.1-5 Subcomponents Within the Scope of Renewal ReviewSFA ......................................2-9 2.1-6 Subcomponents Not Within the Scope of Renewal ReviewTSC................................2-9 2.1-7 Subcomponents Not Within the Scope of RenewalVCC ..........................................2-10 2.1-8 Subcomponents Not Within the ScopeTFR and Transfer Adapter Plate............. .2-10 3.3-1 AMREnvironments ....................................................................................................3-2 3.3-2 AMR ResultsTSC (YR, CY, and LACBWR) ...............................................................3-4 3.3-3 AMR ResultsVCC (YR, CY, and LACBWR) ..............................................................3-8 3.3-4 AMR ResultsTFR and Transfer Adapter (YR, CY, and LACBWR) ...........................3-14 3.3-5 AMR ResultsSFAs ..................................................................................................3-16 3.5-1 AMP Review ResultsLocalized Corrosion and Stress Corrosion Cracking of Welded Stainless-Steel TSCs .................................................................................3-29 3.5-2 AMP Review ResultsInternal VCCMetallic Components Monitoring ....................3-30 3.5-3 AMP Review ResultsExternal VCCMetallic Components Monitoring ...................3-32 3.5-4 AMP Review ResultsVCC StructuresConcrete Monitoring ...................................3-33 3.5-5 AMP Review ResultsTFRs and Transfer Adapters ..................................................3-34 iii

ABBREVIATIONS AND ACRONYMS ADAMS Agencywide Documents Access and Management System AMID Aging Management Institute of Nuclear Power Operations Database AMP aging management program AMR aging management review ANSI American National Standards Institute ASME American Society of Mechanical Engineers B boron B&PV boiler and pressure vessel BWR boiling-water reactor C Celsius CFR Code of Federal Regulations CoC certificate of compliance CY Connecticut Yankee DFC damaged fuel can EPRI Electric Power Research Institute F Fahrenheit FB fuel basket FSAR final safety analysis report GTCC greater than Class C GWd/MTU gigawatt days per metric ton of uranium INPO Institute of Nuclear Power Operations ISFSI independent spent fuel storage installation ITS important to safety LACBWR Lacrosse Boiling Water Reactor LWR light-water reactor MAPS managing aging processes in storage MPC multipurpose canister NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission PH precipitation hardened PWR pressurized-water reactor RFA reconfigured fuel assembly RTD resistance temperature detector iv

SCC stress corrosion cracking SER safety evaluation report SFA spent fuel assembly SSC structure, system, and component STC storage transport cask TFR transfer cask TLAA time-limited aging analysis TSC transportable storage canister UFSAR updated final safety analysis report VCC vertical concrete cask wt% weight percent YR Yankee Rowe v

INTRODUCTION By letter dated December 18, 2019, as supplemented on August 10, 2021; March 18, 2022; July 22, 2022, and December 21, 2022 the current certificate of compliance (CoC) holder, NAC International, Inc. (hereafter applicant), applied for renewal of CoC No. 1025 for the Multi-Purpose Canister Storage System (hereafter NAC-MPC System) for an additional 40 years beyond the initial certificate period (the period of extended operation) (Agencywide Documents Access and Management System Accession Nos. ML19357A178, ML21231A154, ML22077A831, ML22203A127, and ML22355A119 respectively). This safety evaluation report (SER) generally refers to this application, as supplemented, as the renewal application.

The applicant submitted the renewal application in accordance with the regulatory requirements of Title 10 of the Code of Federal Regulations (10 CFR) 72.240, Conditions for spent fuel storage cask renewal. Because the applicant submitted its renewal request more than 30 days before the CoC expiration date, under 10 CFR 72.240(b), this application constitutes a timely renewal. The applicant documented the technical bases for renewal of the CoC and proposed actions for managing the potential aging effects of the structures, systems, and components (SSCs) of the dry storage system to ensure that they will maintain their intended functions during the period of extended operation.

Under 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, Subpart L, Approval of Spent Fuel Storage Casks, the U.S. Nuclear Regulatory Commission (NRC) approved the NAC-MPC System and issued CoC No. 1025 for 20 years, with effective dates of April 10, 2000 (initial certificate, Amendment No. 0); November 13, 2001 (Amendment No. 1); May 29, 2002 (Amendment No. 2); October 1, 2003 (Amendment No. 3);

October 27, 2004 (Amendment No. 4); July 24, 2007 (Amendment No. 5); October 4, 2010 (Amendment No. 6); and March 4, 2019 (Amendments No. 7 and 8). CoC No. 1025 can be used for dry storage of spent nuclear fuel in an independent spent fuel storage installation (ISFSI) at power reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, under 10 CFR Part 72, Subpart K, General License for Storage of Spent Fuel at Power Reactor Sites.

The NAC-MPC System is provided in three configurations for use at (1) Yankee Atomic Electric Companys Yankee Rowe (YR) Nuclear Station (hereafter YR-MPC), (2) Connecticut Yankee (CY) Haddam Neck Nuclear Power Plant (hereafter CY-MPC), and Dairyland Power Cooperative La Crosse Boiling Water Reactor (LACBWR) Nuclear Power Plant (hereafter LACBWR-MPC). Each NAC-MPC System includes a transportable storage canister (TSC) provided with a fuel basket designed to accommodate the allowable spent fuel contents, a vertical concrete cask (VCC), and a transfer cask (TFR) sized to accommodate the pertinent TSC. The YR-MPC, CY-MPC, and LACBWR-MPC have similar components and operating features but different physical dimensions, weights, fuel contents, and storage capacities. All configurations are designed to allow for subsequent transport of the dry-stored spent fuel contents inside each TSC using the certified NAC Internationals storage transport cask package.

The TSC provides the confinement pressure boundary, heat transfer, criticality control, and structural integrity for the safe dry storage of the spent fuel contents. The TSC is stored in the central cavity of the VCC. The VCC provides radiation shielding and structural protection for the vi

TSC and contains internal air flow paths that allow the decay heat from the TSC contents to be removed by natural air circulation around the TSC shell. The principal components of the NAC-MPC System include the following:

  • TSC (YR-MPC, CY-MPC, and LACBWR-MPC) with pressurized-water reactor or boiling-water reactor fuel basket (and damaged fuel cans)
  • VCC (YR-MPC, CY-MPC, and LACBWR-MPC)
  • TFR (YR-MPC as modified and transferred or sold to LACBWR-MPC, and CY-MPC) and transfer adapter
  • spent fuel assemblies
  • fuel transfer and auxiliary equipment (e.g., lift yoke, vertical cask transporter, air pads, heavy haul transfer trailer, vacuum drying and helium back-fill system with a helium mass spectrometer leak detector, welding equipment)
  • VCC temperature monitoring system
  • ISFSI security equipment In the renewal application, the applicant documented the technical bases for renewal of the CoC and proposed actions for managing potential aging effects on the NAC-MPC System SSCs that are within the scope of CoC renewal to ensure that they will maintain their intended functions during the period of extended operation. The applicant presented general information about the dry storage system design and a scoping evaluation to determine the SSCs within the scope of CoC renewal (the in-scope SSCs) and subject to an aging management review. The applicant further screened the in-scope SSCs to identify and describe the subcomponents that support the intended functions of the in-scope SSCs. For each in-scope SSC subcomponent with an identified aging effect, the applicant proposed an aging management program (AMP) or provided a time-limited aging analysis to give assurance that the SSC will maintain its intended function(s) during the period of extended operation.

The NRC staff reviewed the applicants technical bases for safe operation of the NAC-MPC System for an additional 40 years beyond the current CoC term of 20 years. This SER summarizes the results of the staffs review for compliance with 10 CFR 72.240. In its review of the application and development of the SER, the staff used the guidance in (1) NUREG-1927, Revision 1, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, issued June 2016 (NRC, 2016), and (2) NUREG-2214, Revision 0, Managing Aging Processes In Storage (MAPS) Report, issued July 2019 (NRC, 2019). NUREG-2214 establishes a generic technical basis for the safety review of storage renewal applications, in terms of the evaluation of (1) aging mechanisms and effects that could affect the ability of SSCs to fulfill their safety functions in the period of extended operation (i.e., credible aging mechanisms and effects) and (2) aging management approaches to address credible aging effects, including examples of AMPs that are considered generically acceptable to address the credible aging effects to ensure that the design bases will be maintained in the period of extended operation. The staff evaluated the applicants technical basis for its aging management review and proposed AMPs and compared it to the generic vii

technical basis in NUREG-2214. For this comparison, the staff ensured that the design features, environmental conditions, and operating experience for the NAC-MPC System are bounded by those evaluated in NUREG-2214.

This SER is organized into six sections. Section 1 includes the staffs review of the general information of the dry storage system. Section 2 presents the staffs review of the scoping evaluation performed for determining which SSCs are within the scope of renewal. Section 3 provides the staffs evaluation of the aging management review for the assessment of aging effects and aging management activities for SSCs within the scope of renewal. Section 4 documents the additions and changes to the CoC conditions and technical specifications being made to the initial CoC and associated amendments due to renewal. Section 5 presents the staffs conclusions of the safety review. Section 6 lists the references supporting the staffs review and technical determinations.

viii

1 GENERAL INFORMATION 1.1 Certificate of Compliance and Certificate of Compliance Holder Information On December 18, 2019, as supplemented on August 10, 2021; March 18, 2022; July 22, 2022, and December 21, 2022 (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML19357A178, ML21231A154, ML22077A832, ML22203A127, and ML22355A119, respectively), NAC International, Inc. (hereafter NAC or applicant), submitted an application to renew Certificate of Compliance (CoC) No. 1025 for the Multi-Purpose Canister Storage System (hereafter NAC-MPC System). The application is subject to the provisions of Title 10 of the Code of Federal Regulations (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, Subpart K, General License for Storage of Spent Fuel at Power Reactor Sites, and Subpart L, Approval of Spent Fuel Storage Casks.

The applicant requested renewal of the initial NAC-MPC System CoC and Amendments 1 through 8. The U.S. Nuclear Regulatory Commission (NRC) issued the initial CoC (Amendment 0) for the NAC-MPC System on April 10, 2000. Subsequently, the NRC issued eight amendments (1 through 8) to the NAC-MPC System CoC. In its renewal application, NAC described the licensing basis for the NAC-MPC System initial issuance, as well as general descriptions of the changes and reasons for each amendment, including the dates of the applications and associated supplements, the date of issue of the CoC and CoC amendments, and the corresponding updated final safety analysis report (UFSAR) revisions that incorporated the changes. Chapter 1 of the application lists the amendments and describes each amendment along with references. Chapter 2 of the application provides further details on the scope of each amendment.

1.2 Safety Review The objective of this safety review is to determine whether the dry storage system will continue to meet the requirements of 10 CFR Part 72 for an additional 40 years beyond the initial certificate period. The NRC staff safety review is a detailed and in-depth assessment of the technical aspects of the NAC-MPC System renewal application. Pursuant to 10 CFR 72.240(c)(2) and 10 CFR 72.240(c)(3), an application for renewal of a CoC must be accompanied by a safety analysis report that includes (1) time-limited aging analyses (TLAAs) to demonstrate that structures, systems, and components (SSCs) important to safety (ITS) will continue to perform their intended functions for the requested period of extended operation and (2) a description of the aging management programs (AMPs) for management of issues associated with aging that could adversely affect ITS SSCs.

The applicant stated that the renewal application includes the information required by 10 CFR 72.240(c), and the application content is based on the guidance provided in NUREG-1927, revision 1, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, issued June 2016 (NRC, 2016). The applicant also referenced NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, issued July 2019 (NRC, 2019); Nuclear Energy Institute (NEI) 14-03, Revision 2, Guidance for Operations-Based Aging Management for Dry Cask Storage, issued 2016 (NEI, 2016); and other technical references used in support of the renewal application.

1-1

The applicant conducted a scoping evaluation and aging management review (AMR) to identify all SSCs within the scope of the CoC renewal and pertinent aging mechanisms and effects, respectively. The applicant developed AMPs and evaluated TLAAs to ensure that the SSCs identified to be within the scope of renewal will continue to perform their intended functions during the period of extended operation. This safety review documents the staffs evaluation of the applicants scoping evaluation, AMR, and supporting AMPs and TLAAs.

1.3 Application Content The applicants license renewal application contained the following:

  • general information
  • scoping evaluation
  • proposed changes to CoC No. 1025 and technical specifications
  • preapplication inspection report
  • design-basis document review report The applicant also provided UFSAR revisions for all CoC amendments, which incorporated all changes to the NAC-MPC System previously made without prior NRC approval in accordance with 10 CFR 72.48(c) and (d). The UFSAR supplement and changes document provided in Attachment A of the application details the changes and additions for the NAC-MPC System. In addition, during the staffs review of the renewal application, the applicant submitted its 2020 biennial update to the UFSAR (ML20108E871). The staff considered these additional UFSAR revisions in its review of the applicants scoping evaluation and AMR.

1.4 Evaluation Findings

The staff reviewed the general information in the license renewal application, following the guidance in NUREG-1927. Based on its review, the staff determined that the applicant has provided sufficient information with adequate details to support the license renewal application, with the following findings:

F1.1 The information presented in the renewal application satisfies the requirements of 10 CFR 72.240, Conditions for spent fuel storage cask renewal.

F1.2 The applicant tabulated all supporting information and docketed material incorporated by reference, in compliance with 10 CFR 72.240.

1-2

2 SCOPING EVALUATION As described in NUREG-1927, a scoping evaluation is necessary to identify the SSCs requiring an AMR. The objective of this scoping evaluation is to identify SSCs meeting the following criteria:

(1) SSCs that are classified as ITS, as they are relied on for one of the following functions:

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

- Prevent damage to the spent fuel during handling and storage.

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

(2) SSCs that are classified as not ITS but, according to the design bases, the failure of which could prevent fulfillment of a function that is ITS After the determination of in-scope SSCs, the SSCs are screened to identify and describe the subcomponents that support the SSC intended functions.

2.1 Scoping and Screening Methodology In section 2 of the renewal application, the applicant performed a scoping evaluation and provided the following information:

  • a description of the scoping and screening methodology for the inclusion of SSCs and SSC subcomponents in the scope of renewal review
  • a list of sources of information used for the scoping evaluation
  • descriptions of the SSCs
  • a list of the SSCs identified to be within and outside the scope of renewal review and the basis for the scope determination The staff reviewed the scoping process and results provided in the renewal application. The following section discusses the staffs review and findings regarding the applicants scoping evaluation.

2.1.1 Scoping Process In section 2 of the renewal application, the applicant reviewed the following design-bases documents to identify SSCs with safety functions meeting either scoping criterion 1 or 2, as defined above:

  • NAC International, Inc., Final Safety Analysis Report (FSAR) for the NAC-MPC Multi-Purpose Canister System, Docket No. 72-1025 2-1

- NAC-MPC System FSAR, Revision 0, issued May 2000

- NAC-MPC System FSAR, Revision 1, issued February 2002

- NAC-MPC System FSAR, Revision 2, issued November 2002

- NAC-MPC System FSAR, Revision 3, issued March 2004

- NAC-MPC System FSAR, Revision 4, issued November 2004

- NAC-MPC System FSAR, Revision 5, issued October 2005

- NAC-MPC System FSAR, Revision 6, issued November 2006

- NAC-MPC System FSAR, Revision 7, issued November 2008

- NAC-MPC System FSAR, Revision 8, issued February 2009

- NAC-MPC System FSAR, Revision 9. issued November 2010

- NAC-MPC System FSAR, Revision 10, issued January 2014

- NAC-MPC System FSAR, Revision 11, issued April 2018

- NAC-MPC System FSAR, Revision 12, issued April 2019

  • CoC 1025 for the original certificate and the approved amendments:

- NAC-MPC CoC; Initial Issue Amendment 0, Effective date: April 10, 2000

- NAC-MPC CoC; Amendment No. 1, Effective date: November 13, 2001

- NAC-MPC CoC; Amendment No. 2, Effective date: May 29, 2002

- NAC-MPC CoC; Amendment No. 3, Effective date: October 1, 2003

- NAC-MPC CoC; Amendment No. 4, Effective date: October 27, 2004

- NAC-MPC CoC; Amendment No. 5, Effective date: July 24, 2007

- NAC-MPC CoC; Amendment No. 6, Effective date: October 4, 2010

- NAC-MPC CoC; Amendment No. 7, Effective date: March 4, 2019

- NAC-MPC CoC; Amendment No. 8, Effective date: March 4, 2019

  • NAC International, Inc., Calculation No. 455-9000, R0, NAC-MPC Certificates of Compliance Amendment Reconciliation for the Fabrication & Construction of Yankee MPC Transportable Storage Canisters [TSCs], Vertical Concrete Casks [VCCs],

Operational Procedures, and Fuel Contents, dated January 15, 2010

  • NAC International, Inc., Supplemental Certificate of Conformance YR-COC-TSC 1-15NCC 1-15/DFC [Damaged Fuel Can] 1-11, Yankee Atomic Power Company, dated January 22, 2010
  • NAC International, Inc., Calculation No. 12414-9000, R0, Connecticut Yankee Atomic Power Company ISFSI [Independent Spent Fuel Storage Installation] Spent Fuel Storage Project, NAC-MPC Certificate of Compliance Amendment Reconciliation for the Fabrication & Construction of MPC Transportable Storage Canisters, Vertical Concrete Casks and Transfer Casks [TFRs], Operational Procedures, and Fuel Contents, dated January 15, 2010
  • NAC International, Inc., Supplemental Certificate of Conformance CY-COC-TSC-VCC-DFC-TFR for Connecticut Yankee Atomic Power Company, dated January 22, 2010 The applicants scoping process identified SSCs as being either scoped into the review under scoping criteria 1 and 2 described above or not scoped into the review for items not ITS that did not meet scoping criterion 2.

2-2

The staff reviewed the applicants scoping process and determined that it was acceptable because the applicant evaluated the scope of items in the renewal review in a manner that is consistent with NUREG-1927, section 2.4.

2.1.2 Scoping Results Safety evaluation report (SER) table 2.1-1 lists the SSCs the applicant included and excluded from the scope of renewal review and identifies the scoping criterion met by each in-scope SSC.

Table 2.1-1 SSCs Within and Not Within the Scope of Renewal Review SSCs Criterion 11 Criterion 22 In-Scope TSC Yes N/A Yes VCC Yes N/A Yes TFR and Transfer Adapter Yes N/A Yes3 SFAs4 Yes N/A Yes Fuel Transfer and Auxiliary Operating No No No Equipment VCC Temperature Monitoring System No No No ISFSI Storage Pad No5 No No5 ISFSI Security Equipment No No No 1 SSC is ITS.

2 SSC is not ITS, but its failure could prevent fulfillment of an ITS function.

3 Applicable to sites that still retain a TFR or transfer adapter on site, and to TFRs in storage under NAC control; not applicable to facilities that have disposed of the equipment or the equipment is no longer available.

4 Fuel pellets are not included.

5 See further discussion below. The applicant stated that the ISFSI storage pad is not an ITS component within the scope of the CoC renewal; however, general licensees may have designated their pads as ITS. Table 2.3-1 of the application labeled the scoping of the pad as yes (in-scope) to reflect those site-specific cases.

The staff reviewed the scoping results to determine whether the applicant included all SSCs in the approved design bases and whether the conclusions on the out-of-scope SSCs accurately reflect the design-bases documentation. The staff concluded the following on the SSCs excluded from the scope of renewal review:

  • Fuel Transfer and Auxiliary Operating Equipment The applicant stated that the fuel transfer and auxiliary operating equipment necessary for ISFSI operations (e.g., hardware to position the TFR with respect to the storage cask, lifting yoke, lifting slings, air pallets, heavy haul trailer, vertical cask transporter, suction pump equipment, vacuum drying system, welding equipment, weld inspection equipment, and helium backfill and leak detection equipment) are not part of the approved NAC-MPC System under CoC No. 1025 and are not described in detail in the NAC-MPC System FSARs. The applicant also stated that failure of this equipment would not prevent the TSC, VCC, TFR, or spent nuclear fuel assemblies (SFAs) from fulfilling 2-3

their intended safety functions. Therefore, the applicant concluded that the fuel transfer and auxiliary operating equipment do not meet scoping criterion 2 and are not within the scope of renewal review.

The applicant further explained that a majority of this equipment was disposed of following completion of the spent fuel loading operations and decommissioning of the reactor plants. The applicant stated that, when required for operations for removing the loaded NAC-MPC TSCs from the ISFSls, new or refurbished equipment will be provided to complete the fuel transfer operations.

The staff notes that the NAC-MPC System CoC specifically states that the subject equipment is not part of the CoC. In addition, the staff reviewed the design bases of the system in the FSAR and did not identify any means by which the failure of the equipment would affect an ITS function of the TSC, VCC, TFR, or SFAs. Therefore, based on its review of the applicants design-basis documentation, the staff finds the applicants determination that the fuel transfer and auxiliary operating equipment is not within the scope of renewal review to be acceptable.

  • VCC Temperature Monitoring System The applicant stated that the VCC temperature monitoring system is one method authorized to verify the continued operability of the VCC heat removal system, although it is not part of the approved design bases under CoC No. 1025 and, as such, is not described in detail in the NAC-MPC System FSARs. The applicant explained that a VCC temperature monitoring system typically uses thermocouples or resistance temperature detectors placed in each of the four outlet vents. As an alternative to the VCC temperature monitoring system, a visual inspection is performed on a 24-hour frequency to verify that the VCC inlet and outlet screens are unobstructed such that passive cooling air flow is maintained through the VCC.

The applicant stated that the failure of the temperature monitoring equipment would not prevent the VCC from maintaining the stored fuel cladding and MPC components within allowable temperature limits for a period exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, allowing corrective actions to be taken to reestablish operability of the VCC heat removal system. Therefore, the applicant concluded that the VCC temperature monitoring system does not meet scoping criterion 2 and is not within the scope of the renewal review.

The staff reviewed the design bases of the system in the FSAR and verified that the temperature monitoring equipment is not identified as an ITS SSC. The staff also determined that a failure of this monitoring equipment would have no direct effect on the ITS functions of the NAC-MPC System because, as an alternative, a visual inspection is performed on a 24-hour frequency to verify that the VCC inlet and outlet screens are unobstructed, such that passive cooling air flow is maintained through the VCC.

Therefore, the staff finds the applicants determination that the VCC temperature monitoring system is not within the scope of renewal review to be acceptable.

  • ISFSI Pad The ISFSI pad provides free-standing support of the NAC-MPC System casks. The applicant stated that the pad is not part of the approved NAC-MPC System under CoC No. 1025. The applicant also stated that both the FSARs and CoCs authorize the 2-4

evaluation of the ISFSI pad on a site-specific basis as part of the evaluation in 10 CFR 72.212, Conditions of general license issued under 10 CFR 72.210. However, if a general licensee has classified the ISFSI pad as ITS in its respective 10 CFR 72.212 evaluation, then aging management activities will be addressed on a site-specific program basis by the general licensee, as required in its 10 CFR 72.212 evaluation.

The staff reviewed the FSAR and confirmed that the ISFSI pad is not included in the description of the NAC-MPC System components contained in FSAR tables 2.3-1 (Yankee Rowe (YR)-MPC), 2.3-2 (Connecticut Yankee (CY)-MPC), and 2A.3-1 (La Crosse Boiling Water Reactor (LACBWR)-MPC). Similarly, the CoC does not include the ISFSI pad in the description of the principal components of the system. Therefore, the staff finds the applicants determination that the ISFSI pad is not within scope of renewal review to be acceptable.

  • ISFSI Security Equipment The applicant stated that the ISFSI security equipment (e.g., ISFSI security fences and gates, lighting, communications, monitoring equipment) are not part of the NAC-MPC System approved under CoC No. 1025 and, as such, the FSAR does not describe them.

The applicant explained that existing plant programs and procedures ensure that the ISFSI security equipment requirements are met in accordance with 10 CFR Part 73, Physical Protection of Plants and Materials. Furthermore, the applicant stated that potential failure of the ISFSI security equipment would not prevent the NAC-MPC System casks from performing their intended functions.

The staff reviewed the FSAR and CoC and confirmed that the ISFSI security equipment is not included as an ITS SSC in the NAC-MPC System. The staff also evaluated the implications of the failure of the security equipment and did not identify any means by which the security equipment failure could affect an ITS function of the storage system.

As a result, NUREG-1927 specifically recommends the exclusion of the ISFSI security equipment from the scope of renewal review. Therefore, based on its review of the applicants design-basis documentation, and consistent with the review guidance in NUREG-1927, the staff finds the applicants determination that the security equipment is not within the scope of renewal review to be acceptable. Notwithstanding this finding, the staff notes that there are requirements for security equipment to be maintained, but these requirements are not associated with the CoC renewal. In accordance with 10 CFR 73.55, Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage, applicants must maintain the performance of security equipment at all times (not just during the period of extended operation).

Based on its review, the staff finds that the applicant has identified the in-scope SSCs in a manner consistent with NUREG-1927, and therefore, the staff finds the scoping results to be acceptable. The applicant screened the in-scope SSCs to identify and describe the subcomponents that support the SSC intended functions. SER section 2.1.3 describes the SSC subcomponents within and outside the scope of renewal review.

2.1.3 Structures, Systems, and Components Within the Scope of Renewal Review 2-5

As discussed above, the applicant identified the TSC, VCC, TFR and transfer adapter plate, and the SFAs to be within the scope of renewal review. These SSCs consist of several subcomponents, not all of which support an intended function and need be considered in the AMR.

The staff reviewed the applicants screening of the in-scope SSCs to identify subcomponents within the scope of the renewal review. The staffs review considered the intended function of the subcomponent; its safety classification or basis for inclusion in, or exclusion from, the scope of renewal review; and design-basis information in the FSAR, as documented in the system drawings and FSAR tables 2.3-1 (YR-MPC), 2.3-2 (CY-MPC), and 2A.3-1 (LACBWR-MPC).

Based on this review, the staff finds that the applicant screened the in-scope SSCs in a manner consistent with NUREG-1927, and, therefore, the staff finds the screening results for in-scope SSC subcomponents to be acceptable. Below, tables 2.1-2 through 2.1-5 tabulate the subcomponents within the scope of renewal review. Tables 2.1-6 through 2.1-8 tabulate the subcomponents not within the scope of renewal review.

Table 2.1-2 Subcomponents Within the Scope of Renewal ReviewTSC YR-MPC CY-MPC LACBWR-MPC

  • Bottom
  • Bottom
  • Bottom Plate
  • Shield Lid Support Ring
  • Shield Lid Support Ring
  • Closure Lid
  • Spacer Ring
  • Spacer Ring
  • Closure Lid Support Ring
  • Shield Lid
  • Shield Lid
  • Inner Port Cover
  • Structural Lid
  • Structural Lid
  • Closure Ring
  • Port Cover
  • Port Cover
  • Spacer
  • Shield LidDamaged Fuel
  • Bolt
  • Neutron Absorber
  • Nord-Lock Washer
  • Neutron Absorber
  • Cladding
  • Outer Port Cover
  • Cladding
  • Bottom FB Plate
  • PWR Oversized Fuel Tube
  • Bottom FB Weldment
  • Bottom FB Plate
  • Neutron Absorber Pad
  • Bottom FB Weldment Pad
  • Cladding
  • Bottom FB Weldment
  • Bottom FB Weldment
  • Tube Flange Support Plate Support Plate
  • Bottom FB Plate
  • Top FB Weldment Plate
  • Bottom Oversized FB Plate
  • Bottom FB Weldment Pad
  • Top FB Weldment Ring
  • Bottom Weldment FB Plate
  • Bottom FB Weldment Support
  • Top FB Weldment Damaged Fuel Plate Support Plate
  • Top FB Plate
  • Top FB Plate
  • Top FB Weldment
  • Top FB Structural Ring
  • Top FB Structural Ring Stiffener-A
  • Top FB Weldment Support
  • Top FB Weldment Support
  • Top FB Weldment Plate Plate Stiffener-B
  • Top FB Oversized Plate
  • Baffle
  • Top FB PlateDamaged
  • FB Shield Baffle
  • Neutron Absorber Fuel
  • FB Support Disk
  • Cladding
  • FB Support Disk
  • FB Bottom Spacer
  • Spacer
  • Top Spacer
  • Plate 2-6

YR-MPC CY-MPC LACBWR-MPC

  • Bottom Spacer
  • Top Nut
  • FB Support Disk
  • Top Nut
  • Tie Rod
  • FB Heat Transfer Disk
  • Tie Rod
  • Split Spacer
  • Spacer
  • Split Spacer
  • Washer
  • Bottom Spacer
  • Top Spacer
  • FB Heat Transfer Disk
  • Top Nut
  • FB Heat Transfer Disk
  • DFC Collar
  • Tie Rods
  • PWR Basket Flat Washer
  • DFC Lid Plate
  • Top Spacer
  • Top Weldment Baffle A
  • Wiper
  • Split Spacer
  • Top Weldment Baffle B
  • DFC Bottom Plate
  • Flat Washer
  • Screen Cover Plate
  • Filter Screen
  • DFC Collar
  • DFC Lid Plate
  • Backing Screen
  • DFC Lid Plate
  • Wiper
  • Side Plate
  • Wiper
  • Lid Bottom Plate
  • DFC Tube Body
  • DFC Bottom Plate
  • Filter Screen
  • Lift Tee
  • Filter Screen
  • Backing Screen
  • Support Ring
  • Backinq Screen
  • DFC Bottom Plate
  • Lid Bottom Plate
  • Side Plate
  • DFC Collar Side Plate
  • DFC Tube Body
  • DFC Tube Body
  • RFA Corner Angle
  • Lift Tee
  • Lift Tee
  • RFA Tube
  • Support Ring
  • Support Ring
  • Filter Screen
  • Lid Bottom Plate
  • Backing Screen
  • Test Assembly Retainer
  • Stand-Off Pin Lower Tab
  • Hex Head Bolt
  • Support Grid
  • Lifting Plate
  • RFA Bottom Housing
  • Gusset
  • Retaining Plate
  • Ring
  • Retaining Ring
  • RFA Shell Casing
  • RFA Top Housing
  • RFA Top Ring
  • Guide Plate
  • RFA Top End Fitting
  • Rod Retaining Plate
  • RFA Cylinder
  • Screen Ring
  • RFA Inside Tab
  • Screen Housing
  • RFA Outside Tab
  • RFA Top End Plate
  • RFA Top End Template
  • RFA Bottom End Fitting
  • RFA Bolt
  • RFA Alignment Pin
  • RFA FB Corner Angle
  • RFA FB Tie Plate
  • RFA FB Fuel Tube
  • RFA FB Top Cap
  • RFA FB Bottom Cap BWR: boiling-water reactor, RFA: reconfigured fuel assembly, PWR: pressurized-water reactor Table 2.1-3 Subcomponents Within the Scope of Renewal ReviewVCC 2-7

YR-MPC CY-MPC LACBWR-MPC

  • VCC Liner Shell
  • VCC Liner Shell
  • VCC Liner Shell
  • Support Ring
  • Support Ring
  • Weldment Bottom Plate
  • Base Weldment Inlet Cover
  • Base Weldment Inlet Cover
  • Inlet Side Plate
  • Base Weldment Shield Ring
  • Base Weldment Shield Ring
  • Inlet Top Plate
  • Base Weldment Bottom Plate
  • Base Weldment Bottom Plate
  • Stand Base Plate
  • Inlet Side
  • Inlet Side
  • Base Plate
  • Inlet Top
  • Inlet Top
  • Nelson Stud
  • Stand Plate
  • Stand Plate
  • Outlet Bottom Plate
  • Baffle Weldment Base Plate
  • Baffle Weldment Base Plate
  • Outlet Top Plate
  • Nelson Stud
  • Nelson Stud
  • Outlet Shield Plate
  • Outlet Bottom
  • Outlet Bottom
  • Outlet Bottom
  • Outlet Top
  • Outlet Top
  • Outlet Top
  • Outlet Shield Plate
  • Outlet Shield Plate
  • Outlet Side
  • Outlet Bottom
  • Outlet Bottom
  • Outlet Back
  • Outlet Top
  • Outlet Top
  • Baffle Weldment
  • Outlet Side
  • Outlet Side
  • Inlet Shield
  • Outlet Back
  • Outlet Back Pipe/Tube/Bar
  • Baffle
  • Baffle Weldment
  • Baffle Cover Plate
  • Lid Bolt
  • Cover
  • Lid Bolt
  • Cover
  • Lid Bolt
  • VCC Lid Bottom Plate
  • VCC Lid
  • VCC Lid
  • Lid Ring
  • Shield Plug Plate
  • Shield Plug Plate
  • VCC Lid Top Plate
  • Neutron Shield Retaining
  • Neutron Shield Retaining Ring
  • Concrete Ring
  • Neutron Shield Cover Plate
  • Center Support
  • Neutron Shielding
  • Neutron Shielding
  • Nelson Stud
  • Neutron Shield Cover Plate
  • VCC Rebar
  • VCC Rebar
  • Rebar
  • Concrete Shell
  • Concrete Shell
  • Concrete Shell
  • VCC Inlet Supplemental Shield Side Plate
  • Shield Pipe Table 2.1-4 Subcomponents Within the Scope of Renewal ReviewTFR and Transfer Adapter Plate YR-MPC and LACBWR-MPC CY-MPC
  • Bottom Plate
  • Bottom Plate
  • Inner Shell
  • Inner Shell
  • Gamma Shield Brick
  • Gamma Shield Brick
  • Outer Shell
  • Outer Shell
  • Trunnion
  • Trunnion
  • Neutron Shield
  • Neutron Shield
  • Top Plate
  • Top Plate
  • Door Rail
  • Door Rail 2-8

YR-MPC and LACBWR-MPC CY-MPC

  • Shield Door A
  • Shield Door A
  • Shield Door B
  • Shield Door B
  • Door Lock Bolt
  • Door Lock Bolt
  • Retaining Ring
  • Retaining Ring
  • Retaining Ring Bolt
  • Retaining Ring Bolt
  • Connector
  • Connector
  • Door Lock Bolt
  • Lock Pin
  • Strut Bracket
  • Transfer Adapter
  • Hex Head Bolt
  • Lock Pin
  • Transfer Adapter Table 2.1-5 Subcomponents Within the Scope of Renewal ReviewSFA
  • Fuel Rod Cladding
  • Guide Tubes (PWR) or Water Channels (BWR)
  • Spacer Grids
  • Lower and Upper End Fittings
  • Fuel Channel (BWR)
  • Poison Rod Assemblies (PWR)

Table 2.1-6 Subcomponents Not Within the Scope of Renewal ReviewTSC YR-MPC CY-MPC LACBWR-MPC

  • Location Lug
  • Location Lug
  • Location Lug
  • Weather-Resistant Paint
  • Weather-Resistant Paint
  • Weather-Resistant Paint (Alignment Mark) on TSC (Alignment Mark) (Alignment Mark) on TSC Shell
  • Key Shell
  • Metal Boss Seal
  • Valved Nipple
  • Nipple
  • Valved Nipple
  • Seal
  • Seal
  • Key
  • Lubricant
  • Key
  • Weather-Resistant Paint
  • Weather-Resistant Paint
  • Closure Lid Plug (Alignment Mark) on (Alignment Mark) on
  • Weather-Resistant Paint Structural Lid Structural Lid (Alignment Mark) on
  • Shield Lid Plug
  • Shield Lid Plug Closure Lid
  • Structural Lid Plug
  • Structural Lid Plug
  • Drain Tube Nipple
  • Drain Tube
  • Valved Nipple
  • Lubricant
  • Seal
  • Tube
  • Valved Nipple
  • Metal Boss Seal
  • Tube
  • Lid Guide
  • Seal
  • Lid Guide
  • Lubricant
  • Lid Guide 2-9

Table 2.1-7 Subcomponents Not Within the Scope of Renewal ReviewVCC YR-MPC CY-MPC LACBWR-MPC

  • Jack Base
  • Jack Base
  • Screen Tab
  • Jack Gusset
  • Jack Gusset
  • Jack Screw
  • Jack Screw
  • Primer and Paint for
  • Jack Nut
  • Jack Nut Liner, Pedestal, and
  • Jam Nut
  • Jam Nut Baseplate Assemblies
  • Square Nut
  • Square Nut
  • Washer
  • Heavy Hex Nut
  • Primer and Paint for VCC
  • Primer and Coating for Liner,
  • Lifting Nut Lid Pedestal, and Baseplate
  • Primer and Paint for Liner,
  • VCC Nameplate Assemblies Pedestal, and Baseplate
  • Black Paint
  • Washer Assemblies
  • RTD Mounting Plate
  • Insulation
  • Security Seal
  • Screen Strips
  • Seal Tape
  • Washer
  • Vent Screen
  • Seal Wire
  • Seal Tape
  • Screen Bolt
  • Security Seal
  • Seal Wire
  • Plain Washer
  • Tab
  • Primer and Paint for VCC Lid
  • Concrete Anchors
  • Coating System for VCC Lid
  • Lifting and Center Boss
  • Cap Screw
  • Coating System for VCC
  • Primer and Paint for Shield
  • Sealer Shield Plug Plug
  • Retainer Plate
  • Screen Strips
  • Vent Screen
  • Inlet Screen
  • Vent Screen
  • Vent Strips
  • Screen Bolt
  • Screen Bolt
  • Screen Bolt
  • Plain Washer
  • Concrete Anchor
  • RTD Connection Head
  • Concrete Anchor
  • Flat Washer
  • Lag Bolt
  • Lag Bolt
  • Sealer
  • Sealer
  • Coating System for VCC
  • Screen Bolt Supplemental Shield
  • Washer
  • Retainer Plate
  • VCC Nameplate
  • Nameplate
  • Black Weather-Resistant Paint RTD: resistance temperature detector Table 2.1-8 Subcomponents Not Within the Scope of Renewal ReviewTFR and Transfer Adapter Plate YR-MPC and LACBWR-MPC CY-MPC
  • Trunnion Cap
  • Trunnion Cap
  • Scuff Plate
  • Scuff Plate
  • Fill/Drain Line Plate
  • Fill/Drain Line Plate
  • Fill/Drain Line Pipe
  • Fill/Drain Line Pipe 2-10

YR-MPC and LACBWR-MPC CY-MPC

  • Spent Fuel Pool Compatible Coating System
  • Spent Fuel Pool Compatible Coating System
  • Lubricant
  • Spent Fuel Pool Compatible Lubricant
  • Lead Wool
  • Black Weather-Resistant Paint (for
  • Black Weather-Resistant Paint (for Component Identification)

Component Identification)

  • Nameplate
  • Nameplate
  • Flat Washer
  • Door Stop
  • Safety Wire
  • Door Stop YR-MPC: Yankee Rowe Multi-Purpose Canister LACBWR-MPC: Lacrosse Boiling Water Reactor Multi-Purpose Canister

2.2 Evaluation Findings

The NRC staff reviewed the scoping evaluation in the renewal application. The staff performed its review following the guidance in NUREG-1927. Based on its review, the staff finds the following:

F2.1 The applicant has identified all ITS SSCs and SSCs the failure of which could prevent an SSC from fulfilling its safety function, in accordance with the requirements of 10 CFR 72.3, Definitions, and 10 CFR 72.236, Specific requirements for spent fuel storage cask approval and fabrication.

F2.2 The justification for SSCs determined not to be within the scope of the renewal is adequate and acceptable.

2-11

3 AGING MANAGEMENT REVIEW 3.1 Review Objective The objective of the staffs evaluation of the applicants AMR is to determine whether the applicant has adequately reviewed applicable materials, environments, and aging mechanisms and effects and has proposed adequate aging management activities for in-scope SSCs. The AMR addresses aging mechanisms and effects that could adversely affect the ability of the SSCs and associated subcomponents to perform their intended functions during the period of extended operation.

3.2 Aging Management Review Process The applicant described its AMR process as consisting of five steps:

(1) Identify the materials and environments for all subcomponents of the in-scope SSCs.

(2) Identify aging effects requiring management.

(3) Identify and evaluate the TLAAs.

(4) Identify AMPs for managing aging effects.

(5) Evaluate fuel retrievability.

The applicant stated that it identifies materials of construction for the in-scope SSCs and their associated subcomponents by reviewing the licensing drawings contained in the NAC-MPC System FSARs. The applicant also stated that it identifies the environments to which the materials are normally exposed based on a review of the latest NAC-MPC System FSARs and plant loading procedures and records. The applicant defines and classifies the environments in accordance with those defined in NUREG-2214.

The staff reviewed the applicants AMR process and finds that it is acceptable because it is consistent with the methodology recommended in NUREG-1927 and is adequate for identifying credible aging effects for the SSCs within the scope of renewal review.

3.3 Aging Management Review Results: Materials, Service Environment, Aging Effects, and Aging Management Activities The staff evaluated the applicants technical basis for its AMR by comparing it to the generic technical basis in NUREG-2214. In this evaluation, the staff verified that the design features, environmental conditions, and operating experience for the NAC-MPC System are bounded by those evaluated in NUREG-2214.

The applicant defined the SSC service environments in section 3.1.2 of the renewal application.

Table 3.3-1 of this SER summarizes these environments and compares them to those evaluated in NUREG-2214. The staff considered this comparison in determining whether the conclusions in NUREG-2214 are consistent with the applicants analysis of the NAC-MPC System.

3-1

Table 3.3-1 AMREnvironments Environment in the Equivalent Renewal Description Environment(s) in Application NUREG-2214 Helium Environment inside the TSC that is vacuum Helium dried and backfilled with inert helium gas and has very low levels of oxygen and moisture.

Fully encasedsteel Environment that applies to materials that Embedded in metal are fully enclosed within another steel component or fully lined by steel, which Fully encased or lined prevents the ingress of water or contaminants.

In addition, this environment includes small free volumes that exist between components but are sealed from replenishment by outside air.

Sheltered Environment that includes ambient air but Sheltered not sunlight, rain, or wind exposure. The ambient air may contain moisture and salinity.

Embedded Environment applicable to metal components Embedded in concrete (concrete) that are cast inside or against concrete.

Airoutdoor Environment for exterior surfaces that are Airoutdoor exposed to all weather-related effects, including sunlight, wind, and rain/snow/ice (with potential for salts), and has temperature ranges equivalent to the site ambient temperature ranges.

Airindoor/outdoor Environment for components that are Airoutdoor typically housed indoors, except for periodic outdoor exposure during transfer operations. (includes the The typical environment may be a building, indoor/outdoor air outdoor storage container, or other environment described by protective covering. the applicant)

In the AMR, the applicant conservatively evaluated these components as being normally exposed to outdoor air.

Tables 3.3-2 through 3.3-5 summarize the results of the applicants AMR and identify the disposition of each potential aging effect for SSC subcomponent materials within the scope of renewal review. These tables identify whether the applicants conclusions regarding the credibility of each aging effect are consistent with the generic technical bases and conclusions 3-2

in NUREG-2214. The tables also identify the disposition of the aging effect in terms of whether (1) an aging management activity (i.e., AMP or TLAA) is, or is not, needed to address the aging effect (consistent with NUREG-2214) or (2) there is a separate technical basis or supporting analysis that justifies either that an aging effect is not credible or that an aging management activity is not needed for the aging effect (for items either not addressed in, or inconsistent with, NUREG-2214).

3-3

Table 3.3-2 AMR ResultsTSC (YR, CY, and LACBWR)

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Change in AMP/TLAA not Creep No Yes dimensions necessary Fatigue TLAA Not evaluated in Fatigue Cracking Yes (see SER NUREG-2214 section 3.4.1)

Yes, for non-Helium precipitation- AMP/TLAA not Loss of hardened alloys necessary fracture Thermal aging toughness/ No Not consistent Stainless steel loss for 17-4 See SER (austenitic) of ductility precipitation- section 3.3.1.1 hardened alloys Loss of AMP/TLAA not Stress relaxation No Yes preload necessary Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Fatigue TLAA Sheltered Not evaluated in Fatigue Cracking Yes (see SER NUREG-2214 section 3.4.1)

Loss of TSC Localized Pitting and material Corrosion and SCC Yes Yes crevice corrosion (precursor to AMP SCC) (see SER table 3.5-1) 3-4

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 TSC Localized Corrosion and SCC SCC Cracking Yes1 Yes AMP Sheltered (see SER table 3.5-1)

Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Change in AMP/TLAA not Creep No Yes dimensions necessary Fatigue TLAA Not evaluated in Fatigue Cracking Yes (see SER Stainless steel NUREG-2214 section 3.4.1)

(austenitic)

Pitting and Loss of AMP/TLAA not No Yes crevice corrosion material necessary Fully encased AMP/TLAA not SCC Cracking No Yes necessary Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Loss of fracture AMP/TLAA not Thermal aging toughness/ No Yes necessary loss of ductility Steel Helium Radiation AMP/TLAA not Cracking No Yes embrittlement necessary 3-5

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 Change in AMP/TLAA not Creep No Yes dimensions necessary Fatigue TLAA Not evaluated in Fatigue Cracking Yes (see SER NUREG-2214 section 3.4.1)

Loss of AMP/TLAA not Steel Helium General corrosion No Yes material necessary Loss of fracture AMP/TLAA not Thermal aging toughness/ No Yes necessary loss of ductility Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Change in AMP/TLAA not Creep No Yes2 dimensions necessary Aluminum Helium Loss of AMP/TLAA not General corrosion No Yes material necessary Loss of fracture AMP/TLAA not Thermal aging toughness/ No Yes2 necessary loss of ductility Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Neutron absorber Change in AMP/TLAA not (Boral') Helium Creep No Yes dimensions necessary Loss of AMP/TLAA not General corrosion No Yes material necessary 3-6

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 Galvanic Loss of AMP/TLAA not No Yes corrosion material necessary Loss of AMP/TLAA not Thermal aging No Yes strength necessary Neutron absorber Wet corrosion and Change in AMP/TLAA not (BoralTM) Helium No Yes blistering dimensions necessary Neutron Absorber Loss of Not evaluated in and Shield TLAA Boron depletion criticality Yes NUREG-2214 (see SER control section 3.4.3) 1 The applicant concluded that stress corrosion cracking of the stainless steel TSC material exposed to the sheltered air environment is applicable only to welded stainless steel (where sufficient residual stress exists). The staff verified that this conclusion is consistent with NUREG-2214, table 3-2.

2 In section 3.2.1.3 of the renewal application, the applicant concluded that creep and thermal aging are not credible for the aluminum heat transfer discs that do not bear any load beyond their own weight. The staff verified that this conclusion is consistent with NUREG-2214 for nonstructural aluminum components.

3-7

Table 3.3-3 AMR ResultsVCC (YR, CY, and LACBWR)

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Corrosion TLAA (see SER section 3.4.2)

Loss of General corrosion Yes Yes material Internal VCC Metallic Monitoring AMP (see SER table 3.5-2)

Corrosion TLAA Sheltered (see SER section 3.4.2)

Pitting and Loss of Steel Yes Yes crevice corrosion material Internal VCC Metallic Monitoring AMP (see SER table 3.5-2)

Internal VCC Metallic Galvanic Loss of Yes1 Yes Monitoring AMP corrosion material (see SER table 3.5-2)

Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Corrosion TLAA Loss of General corrosion Yes Yes (see SER Embedded material section 3.4.2)2 (Concrete) Corrosion TLAA Pitting and Loss of Yes Yes (see SER crevice corrosion material section 3.4.2)2 3-8

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Loss of Corrosion TLAA concrete/ Yes Yes (see SER steel bond section 3.4.2)2 Corrosion TLAA Embedded Loss of Yes Yes (see SER (Concrete) material Corrosion of section 3.4.2)2 reinforcing steel Corrosion TLAA Cracking Yes Yes (see SER section 3.4.2)2 Corrosion TLAA Loss of Steel Yes Yes (see SER strength section 3.4.2)2 Radiation AMP/TLAA not Cracking No Yes embrittlement necessary External VCC Loss of Metallic Monitoring General corrosion Yes Yes material AMP (see SER table 3.5-3)

External VCC Pitting and Loss of Metallic Monitoring Yes Yes AirOutdoor crevice corrosion material AMP (see SER table 3.5-3)

External VCC Galvanic Loss of Metallic Monitoring Yes Yes corrosion material AMP (see SER table 3.5-3) 3-9

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion AMP/TLAA not SCC Cracking No Yes3 necessary Stainless steel Radiation AMP/TLAA not Sheltered Cracking No Yes embrittlement necessary Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Loss of AMP/TLAA not Stress relaxation No Yes preload necessary Internal VCC Metallic Galvanic Loss of Stainless steel Yes1 Yes Monitoring AMP Sheltered corrosion material (see SER table 3.5-2)

Pitting and Loss of AMP/TLAA not No Yes crevice corrosion material necessary Neutron Absorber Loss of Cement-based Radiation and Shielding TLAA Fully encased shielding Yes No4 neutron shielding damage (see SER effectiveness section 3.4.3)

Neutron Absorber Loss of Polymer-based Radiation and Shielding TLAA Fully encased shielding Yes Yes neutron shielding damage (see SER effectiveness section 3.4.3) 3-10

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Neutron Absorber Loss of and Shielding TLAA Thermal aging shielding Yes Yes (see SER effectiveness section 3.4.3)

Neutron Absorber Loss of and Shielding TLAA Boron depletion5 shielding Yes Yes (see SER effectiveness section 3.4.3)

Reinforced VCC Concrete Airoutdoor Reaction with Cracking Yes Yes Structures AMP aggregates (see SER table 3.5-4)

Reaction with Reinforced VCC aggregates Loss of Yes Yes Structures AMP strength (see SER table 3.5-4)

Loss of Reinforced VCC material Salt scaling Yes Yes Structures AMP (spalling, (see SER table 3.5-4) scaling)

Reinforced VCC Cracking Yes Yes Structures AMP (see SER table 3.5-4)

Reinforced VCC Aggressive Loss of Yes Yes Structures AMP chemical strength (see SER table 3.5-4) attack Loss of Reinforced VCC Concrete Airoutdoor material Yes Yes Structures AMP (spalling, (see SER table 3.5-4) scaling)

AMP/TLAA not Creep Cracking No Yes necessary AMP/TLAA not Shrinkage Cracking No Yes necessary 3-11

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Loss of AMP/TLAA not No Yes Dehydration at strength necessary high temperatures AMP/TLAA not Cracking No Yes necessary AMP/TLAA not Fatigue Cracking No Yes necessary AMP/TLAA not Cracking No Yes Delayed ettringite necessary formation Loss of AMP/TLAA not No Yes strength necessary Delayed ettringite Loss of formation material AMP/TLAA not No Yes (spalling, necessary scaling)

Reinforced VCC Cracking Yes Yes Structures AMP Freezethaw (see SER table 3.5-4)

(above freeze Loss of line) Reinforced VCC material Yes Yes Structures AMP Concrete Airoutdoor (spalling, (see SER table 3.5-4) scaling)

AMP/TLAA not Cracking No Yes Radiation necessary damage Loss of AMP/TLAA not No Yes strength necessary Leaching of Reinforced VCC Loss of calcium Yes Yes Structures AMP strength hydroxide (see SER table 3.5-4) 3-12

Consistent with Aging Applicant Defined Material Environment Aging Effect Conclusion Disposition Mechanism as Credible of NUREG-2214 Increase in Reinforced VCC porosity and Yes Yes Structures AMP permeability (see SER table 3.5-4)

Reduction of concrete pH (reducing Reinforced VCC corrosion Yes Yes Structures AMP resistance of (see SER table 3.5-4) steel embedments) 1 The applicant identified galvanic corrosion as credible only for steel and stainless steel components with a dissimilar metal contact. The staff confirmed that this is consistent with the guidance in NUREG 2214.

2 Although the applicant stated that corrosion of reinforcing steel and other steel components embedded in the VCC concrete is addressed by its corrosion TLAA, the staff determined that the applicants reinforced VCC structures AMP includes inspections for this aging effect that are consistent with the recommendations in NUREG-2214 for adequately managing this aging effect.

3 According to the applicant, the VCC stainless steel components do not contain welds and are not susceptible to stress corrosion cracking (SCC). The staff confirmed that the VCC does not have any welded stainless steel components. Since there are no welded stainless steel VCC components, and therefore no components with significant residual stress, the staff confirmed that there are no VCC stainless steel components that may be susceptible to SCC.

4 NUREG-2214 concluded that radiation damage of cement-based neutron shielding is not credible. However, the applicant considered its radiation exposure test data for cement-based neutron shielding and provided a TLAA that applied this test data to address the potential for radiation damage for 60 years. The staffs evaluation of the applicants TLAA of radiation damage of the VCC cement-based neutron shielding is addressed below in section 3.4.3 of this SER.

5 The staff verified that the applicants analysis of boron depletion was appropriately limited to the polymeric shielding material because the presence of boron is not relied on for shielding in cement-based material.

3-13

Table 3.3-4 AMR ResultsTFR and Transfer Adapter (YR, CY, and LACBWR)

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Loss of Transfer Cask AMP General corrosion Yes Yes material (see SER table 3.5-5)

Pitting and Loss of Transfer Cask AMP Yes Yes crevice corrosion material (see SER table 3.5-5)

Steel Airoutdoor Galvanic Loss of Transfer Cask AMP Yes1 Yes corrosion material (see SER table 3.5-5)

Microbiologically Loss of AMP/TLAA not influenced No Yes material necessary corrosion Loss of Transfer Cask AMP Wear Yes2 Yes material (see SER table 3.5-5)

Radiation AMP/TLAA not Cracking No Yes embrittlement necessary Pitting and Loss of AMP/TLAA not No Yes3 crevice corrosion material necessary Stainless steel Microbiologically Airoutdoor Loss of AMP/TLAA not (austenitic and ferritic) influenced No Yes material necessary corrosion AMP/TLAA not SCC Cracking No Yes3 necessary Loss of AMP/TLAA not Stress relaxation No Yes preload necessary Stainless steel (austenitic and ferritic) Galvanic Loss of Transfer Cask AMP Airoutdoor Yes1 Yes corrosion material (see SER table 3.5-5) 3-14

Applicant Consistent with Aging Material Environment Aging Effect Defined as Conclusion of Disposition Mechanism Credible NUREG-2214 Neutron Absorber and Loss of Shielding TLAA Radiation damage shielding Yes Yes (see SER effectiveness section 3.4.3)

Neutron Absorber and Loss of Polymer-based Shielding TLAA Fully encased Thermal aging shielding Yes Yes neutron shielding (see SER effectiveness section 3.4.3)

Neutron Absorber and Loss of Shielding TLAA Boron depletion shielding Yes Yes (see SER effectiveness section 3.4.3)

AMP/TLAA not Lead Fully encased None None No Yes necessary 1 The applicant identified galvanic corrosion as credible only for components with a dissimilar metal contact. The staff confirmed that this is consistent with the guidance in NUREG 2214.

2 The applicant only identified wear as credible for those components with sliding contact (e.g., trunnions, door rails). The staff confirmed that this is consistent with the guidance in NUREG 2214.

3 The applicant noted that the stainless steel components do not contain welds that may be susceptible to SCC, and pitting and crevice corrosion are only credible aging mechanisms as a precursor to SCC. The staff confirmed that this is consistent with NUREG-2214, table 3-2.

3-15

Table 3.3-5 AMR ResultsSFAs Consistent Applicant Aging with Material Environment Aging Effect Defined as Disposition Mechanism Conclusion of Credible NUREG-2214 Loss of load AMP/TLAA not Oxidation bearing No Yes necessary capacity Loss of AMP/TLAA not Pitting corrosion No Yes material necessary Galvanic Loss of AMP/TLAA not No Yes corrosion material necessary AMP/TLAA not SCC Cracking No Yes necessary Hydride-induced Loss of AMP/TLAA not No Yes Cladding: embrittlement ductility necessary Helium zirconium-based alloy Delayed hydride AMP/TLAA not Cracking No Yes cracking necessary Low-temperature Change in AMP/TLAA not No Yes creep dimensions necessary Radiation Loss of AMP/TLAA not No Yes embrittlement strength necessary AMP/TLAA not Fatigue Cracking No Yes necessary Mechanical AMP/TLAA not Cracking No Yes overload necessary 3-16

Consistent Applicant Aging with Material Environment Aging Effect Defined as Disposition Mechanism Conclusion of Credible NUREG-2214 AMP/TLAA not Not evaluated Loss of necessary General corrosion No in material (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Not evaluated necessary SCC Cracking No in (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Not evaluated Loss of necessary Pitting corrosion No in material (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Not evaluated necessary Stress rupture Cracking No in (see SER NUREG-2214 section 3.3.1.2)

Cladding:

Helium AMP/TLAA not stainless steel alloy Not evaluated Strain rate necessary Cracking No in embrittlement (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Not evaluated Hydrogen-induced necessary Cracking No in degradation (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Not evaluated Helium necessary Cracking No in embrittlement (see SER NUREG-2214 section 3.3.1.2)

AMP/TLAA not Fission product Not evaluated necessary cladding Cracking No in (see SER interaction NUREG-2214 section 3.3.1.2) 3-17

Consistent Applicant Aging with Material Environment Aging Effect Defined as Disposition Mechanism Conclusion of Credible NUREG-2214 Cladding Fatigue Not evaluated TLAA Fatigue Cracking No in (see SER NUREG-2214 section 3.4.4)

Change in AMP/TLAA not Creep No Yes dimensions necessary Guide tubes (PWR) or water channels (BWR); Change in AMP/TLAA not Hydriding No Yes fuel channels: Dimensions necessary Helium zirconium-based alloy Radiation Loss of AMP/TLAA not or No Yes embrittlement strength necessary stainless steel AMP/TLAA not Fatigue Cracking No Yes necessary Change in AMP/TLAA not Creep No Yes dimensions necessary Change in AMP/TLAA not Hydriding No Yes Spacer grids: dimensions necessary Helium zirconium-based alloy Radiation Loss of AMP/TLAA not No Yes embrittlement strength necessary AMP/TLAA not Fatigue Cracking No Yes necessary Change in AMP/TLAA not Creep No Yes dimensions necessary Loss of AMP/TLAA not General corrosion No Yes material necessary Spacer grids; lower and Helium upper end fittings: AMP/TLAA not SCC Cracking No Yes Inconel necessary 3-18

Consistent Applicant Aging with Material Environment Aging Effect Defined as Disposition Mechanism Conclusion of Credible NUREG-2214 Radiation Loss of AMP/TLAA not No Yes embrittlement strength necessary AMP/TLAA not Fatigue Cracking No Yes necessary Change in AMP/TLAA not Creep No Yes dimensions necessary Loss of AMP/TLAA not General corrosion No Yes material necessary Lower and upper end fittings; poison rod AMP/TLAA not Helium SCC Cracking No Yes assemblies (PWR): necessary stainless steel Radiation Loss of AMP/TLAA not No Yes embrittlement strength necessary AMP/TLAA not Fatigue Cracking No Yes necessary 3-19

The staff reviewed the applicants AMR results for consistency with the technical bases for aging mechanisms and effects in NUREG-2214. If it determined that the applicants conclusions were consistent with expected aging management activities in accordance with NUREG-2214, the staff considered the results acceptable and provides no additional discussion in this SER.

The following sections address the staffs review of the applicants conclusions on aging mechanisms and effects for which the staff was not able to verify consistency with NUREG-2214 or for which it considered an additional explanation to be warranted.

3.3.1 Supplemental Analyses The assessments below document the staffs review for those AMR conclusions in the tables above that were either inconsistent with NUREG-2214, were not evaluated in NUREG-2214, or warranted additional explanation.

3.3.1.1 Thermal Aging of 17-4 Precipitation-Hardened Stainless Steel Fuel Basket Support Disks In table 3.2-1 of the renewal application, the applicant did not include an AMR line item to address the potential for thermal aging of the alloy 17-4 precipitation-hardened (PH) stainless steel fuel basket (FB) support disks. NUREG-2214 states that this material may be susceptible to thermal embrittlement above 240 degrees Celsius (°C) (470 degrees Fahrenheit (°F), and applicants should provide an analysis on a case-by-case basis to demonstrate that components constructed of this material do not undergo unacceptable reductions in fracture toughness.

NUREG-2214 references Olender et al. (2015) for guidance on how 17-4 PH alloys should be evaluated with respect to resistance to thermal aging.

In section 3.2.1.2.8 of the renewal application, the applicant stated that the maximum long-term service temperature of the support disks is 316°C (601°F), which is below the 343°C (650°F) maximum allowable service temperature in American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV) Code,Section II, Part D. The applicant also stated that the average support disk temperature is 181°C (358°F). Additionally, in a related request to renew the NAC-Universal Storage System CoC No. 1015, the applicant provided further information on the potential for thermal aging of 17-4 PH basket support disks (ML22062A764). In that correspondence, the applicant noted that most of the operating experience of failures of 17-4 PH components occurred in active reactor components subject to dynamic loading and also involved an element of corrosion, SCC, or an overloading event. The applicant stated that the subject fuel basket support disks are not subject to dynamic loading, the applied static stresses are minimal during normal storage operations, and the disks are within an inert helium service environment that would preclude corrosion-related effects.

In its independent analysis of this aging effect, the staff noted that the drawings show that the support disk alloys conform to ASME SA-693, Type 630, and that the material is heat treated to condition H1150, which refers to the temperature 621°C (1,150°F) at which the alloy is heat treated to promote precipitation hardening. Olender et al. (2015) specifically recommends this heat treatment as optimal to reduce strength and hardness, which, in turn, is expected to reduce susceptibility to fracture. The staff also acknowledges the relatively benign normal service conditions noted by the applicant (i.e., low static stresses, inert helium environment) and the fact that only a portion of the disks experience service temperatures above the 240°C (470°F) embrittlement threshold. The staff also considered the impact performance of the disks during the postulated cask tip-over event. FSAR section 11.2.12 describes the analysis of the stresses in the fuel basket support disks in the tip-over event, including the margins to the allowable 3-20

stresses for various portions of the support disks. The staff notes that no locations of the disks experience overloading (i.e., stresses above the materials yield strength) and those portions of the disks that have the lowest stress safety margin are generally associated with the outer locations of the disks (away from the basket center) that are not expected to reach temperatures above the embrittlement threshold.

As a result, based on (1) the optimal alloy heat treatment, (2) the relatively benign normal service conditions, and (3) the fact that the lowest tip-over safety margins occur in those portions of the basket that are not expected to exceed the embrittlement threshold, the staff finds the applicants determination that the disks are not subject to an unacceptable loss of fracture toughness due to thermal aging to be acceptable.

3.3.1.2 Aging of Stainless Steel Fuel Rod Cladding In section 3.2.5.2 and table 3.2-9 of the renewal application, the applicant concluded that stainless steel fuel rod cladding does not have any aging effects requiring management. In support of its conclusion, the applicant evaluated several potential aging mechanisms, including general corrosion, SCC, localized corrosion (pitting), stress rupture, strain rate embrittlement, hydrogen-induced degradation, helium embrittlement, and fission product cladding interaction.

The applicant separately evaluated the potential for fatigue in a TLAA, as documented in the staffs review of that aging mechanism in SER section 3.4.4. NUREG-2214 does not include a generic evaluation of the aging of stainless steel cladding.

The staff notes that the NRC previously evaluated the potential for degradation of stainless steel cladding in its approval of the initial 20-year CoC for the NAC-MPC System (NRC, 2000). That evaluation was based, in part, on Electric Power Research Institute (EPRI) Technical Report 106440, Evaluation of Expected Behavior of LWR Stainless Steel-Clad Fuel in Long-Term Dry Storage, issued 1996 (EPRI, 1996). In support of the staffs original conclusion that cladding would remain intact for the initial 20-year license term, the staff verified that maximum temperatures of the stainless steel cladding are below temperatures that could lead to stress rupture, creep, sensitization, helium embrittlement, or strain rate embrittlement. In addition, the staffs original conclusion noted that the inert helium environment precludes the corrosion-related aging mechanisms (general corrosion, localized corrosion, and stress corrosion cracking).

In its review of the applicants conclusion that the stainless steel cladding is not subject to aging in the renewed storage term, the staff notes that the extension of storage to 60 years does not alter the premise of the staffs original conclusions in the initial CoC. In the renewed storage term, cladding temperatures are lower than the initial term due to the reduced decay heat, and thus temperatures remain below levels that may cause stress rupture, creep, sensitization, helium embrittlement, or strain rate embrittlement. Also, maintenance of the confinement boundary (through the TSC localized corrosion and SCC AMP evaluated in SER section 3.5) will maintain the inert environment within the canister to prevent corrosion-related aging.

Based on its reviews above, including the evaluation of the effects of the extended storage term on the initial design basis, the staff finds the applicants determination that the stainless steel cladding is not subject to aging to be acceptable.

3.3.2 Evaluation Findings 3-21

The staff reviewed the AMR in the renewal application to verify it adequately identified the materials, environments, and aging effects of the in-scope SSCs. The staff performed its review following the guidance in NUREG-1927 and NUREG-2214. Based on its review of the renewal application, the staff finds the following:

F3.1 The applicants AMR process is comprehensive in identifying the materials of construction and associated operating environmental conditions for those SSCs within the scope of renewal, and the applicant has provided an acceptable summary of the information in the renewal application and the FSAR supplement.

F3.2 The applicants AMR process is comprehensive in identifying all pertinent aging mechanisms and effects applicable to the SSCs within the scope of renewal, and the applicant has provided an acceptable summary of the information in the renewal application and the FSAR supplement.

3.4 Time-Limited Aging Analyses As discussed in section 3.3 of the renewal application, the applicant identified four TLAAs for SSCs within the scope of the renewal review:

(1) fatigue evaluation for NAC-MPC TSC components for extended storage (2) corrosion analysis of NAC-MPC VCC internal steel components for extended storage (3) aging analysis for NAC-MPC neutron absorber and neutron shield components (4) evaluation of stainless-steel clad fuel for fatigue in storage Based on its review of the design-basis documents, the staff confirmed that the applicant identified all calculations and analyses that meet all six criteria in 10 CFR 72.3 that define a TLAA. The following sections document the staffs evaluation of the applicants TLAAs.

3.4.1 Fatigue Evaluation of NAC-MPC TSC Confinement Boundary and Fuel Basket Components for Extended Storage In section 3.3.3.1 of the renewal application, the applicant stated that the TSC confinement boundaries and fuel baskets were determined to meet the ASME B&PV Code fatigue screening criteria for components that do not require further cyclic fatigue analysis. If these fatigue screening conditions are satisfied, the ASME B&PV Code does not require further in-depth analysis for cyclic service and considers that the effects of fatigue will be accounted for by compliance with the other applicable code requirements. Appendix B to the renewal application provides the evaluation details in proprietary Calculation No. 30013-2001, Fatigue Evaluation of MPC and UMS Storage System Components for Extended Storage.

The applicants fatigue screening analysis determined that the NAC-MPC canister confinement boundaries satisfy all conditions specified in the applicable design code, ASME B&PV Code,Section III, NB-3222.4(d)(1) through (6), for not having to perform an in-depth fatigue analysis for a 60-year service life. The applicant also concluded that the fuel baskets satisfy all conditions specified in the applicable design code, ASME B&PV Code,Section III, NG-3222.4(d)(1) through (4), for not having to perform an in-depth fatigue analysis for a 60-year service life. The staff noted that these code conditions are based on a comparison of peak stresses with strain cycling fatigue data and include cyclic stresses generated as a result of (1) atmospheric to service pressure cycles, (2) normal service pressure fluctuation, (3) temperature difference between startup and shutdown, (4) temperature difference in normal 3-22

service, (5) temperature difference between dissimilar metals, and (6) mechanical loads, as applicable to the ASME B&PV Code conditions specified in NB-3222.4(d) and NG-3222.4(d).

The applicant provided an evaluation based on these conditions to show that the ASME B&PV Code fatigue screening requirements are satisfied. Therefore, the applicant concluded that the NAC-MPC canisters and fuel basket components do not require further in-depth fatigue analysis for cyclic service for 60 years of extended storage.

The staff reviewed the applicants fatigue screening calculations for each of the ASME B&PV Code conditions in NB-3222.4(d) and NG-3222.4(d). The staff verified that the applicants calculations adequately demonstrate that all conditions of the ASME B&PV Code are either met or are not applicable; therefore, the staff finds that further detailed fatigue analyses is not required, and the applicants evaluation of the TLAA is acceptable.

3.4.2 Corrosion Analysis of NAC-MPC VCC Internal Steel Components for Extended Storage General Corrosion In section 3.3.3.2 of the renewal application, the applicant analyzed the general corrosion of the VCC internal carbon steel components that are exposed to the sheltered air environment inside the VCC to determine whether corrosion may have an adverse impact on the intended structural, thermal, and radiation shielding safety functions of these components over the 60-year extended storage term. Appendix B to the renewal application provides the evaluation details in proprietary Calculation No. 30013-2003, Corrosion Analysis of MPC VCC Steel Components for Extended Storage.

The applicant selected a constant general corrosion rate of 0.003 inch per year over 60-years as a basis for analyzing the effects of general corrosion on the structural, thermal, and radiation shielding safety functions of the VCC internal carbon steel components. The applicants corrosion calculation determined that the progression of general corrosion at this constant rate over the entire 60-year extended storage term would result in a 0.18 inch reduction in thickness on each side of the exposed surfaces of the VCC internal carbon steel components.

Accordingly, for VCC internal carbon steel items exposed on just one side to the sheltered air environment inside the VCC, the applicant used a total corrosion allowance of 0.18 inch in its analyses; for VCC internal carbon steel items exposed on both sides to the sheltered air environment, the applicant used a total corrosion allowance of 0.36 inch in its analyses. The applicant provided structural, thermal, and radiation shielding analyses to demonstrate that the total corrosion allowance for the internal VCC steel components would not have an adverse effect on the ability of the VCC assembly to perform its intended structural, thermal, and shielding safety functions during the extended 60-year storage term. The applicants analyses determined the following:

  • For the VCC bottom lift by hydraulic jacks, the applicant determined that the maximum bearing stress in the concrete and the maximum stresses in the pedestal with corrosion after a 60-year service life remain within the allowable stress limits. In addition, the applicant determined that the corrosion allowance on the opposite side of the plates to which the nelson studs are welded will not adversely impact the design function of the Nelson studs.
  • The VCC dead load, live load, flood, tornado wind, and seismic loading does not take any structural credit for the VCC steel liner; therefore, the applicant determined that any 3-23

reduction in the VCC liner thickness resulting from corrosion does not change the results of the VCC analysis for these load conditions.

  • The applicant determined that a reduction of the VCC steel liner thickness due to corrosion would result in a negligible change in the thermal stresses in the concrete and rebar.
  • The analysis of local damage to the VCC concrete shell due to tornado missile impacts does not take any structural credit for the VCC steel liner; therefore, the applicant determined that any reduction in the VCC liner thickness resulting from corrosion does not change the results of the VCC analysis for tornado missile impact.
  • The applicant determined that the VCC lid assemblys strength remains adequate to prevent perforation by a missile due to a design-bases tornado, even if its thickness were to be reduced by the corrosion allowance.
  • Regarding the analysis of the VCC 6-inch design-basis drop, the applicant explained that the evaluation of the VCC concrete shell does not take any structural credit for the VCC steel liner; therefore, any reduction in the VCC liner thickness resulting from corrosion does not change the results of the VCC concrete shell for this design-basis drop condition. The applicants structural evaluation of the VCC pedestal concluded that the maximum deformation of the pedestal due to the drop will result in a reduction of the air inlet cross section area; however, this deformation is bounded by the design-basis analysis where the vent inlets are half blocked. The applicant also determined the acceleration experienced by the TSC during this design-basis drop accident if there is corrosion of the VCCs weldment plate and demonstrated that these accelerations will be lower than the accelerations of the TSC with the uncorroded, nominal plate thicknesses pedestal.
  • In a VCC tip-over accident, the applicant determined that general corrosion of the steel inner shell will reduce the overall beam-bending and ring-bending stiffness of the VCC, which will slightly reduce the acceleration loads that are imparted to the TSC and basket components.
  • The applicant determined that corrosion of the steel plates that line the VCC air passage will improve the surface properties with respect to thermal performance, but the expansion of the rust layer into the air passage could reduce the air flow cross section.

The applicant concluded that the net effect of the corrosion of the steel surfaces that line the air passage on the thermal performance of the system is insignificant.

  • The applicant determined that any reduction in gamma shielding resulting from loss of steel due to corrosion over the extended storage period is offset by the radioactive decay of the source over the same timeframe.

In its review of the applicants TLAA, the staff verified that the applicants selected general corrosion rate of 0.003 inch per year for the VCC internal carbon steel components over a 60-year storage period is consistent with data in the technical literature for long-term direct exposure of carbon steels to a marine atmosphere (National Association of Corrosion Engineers, 2016). The staff noted that the applicants selected general corrosion rate and total 60-year corrosion allowance is conservative for this analysis since it does not take credit for the coatings that are present on all exposed steel surfaces, nor does it credit any favorable effect 3-24

associated with elevated temperature inside the VCC that may impede the deliquescence of moisture with dissolved compounds onto the interior surfaces of the VCC; the staff noted that both of these factors provide a significant degree of conservatism for the applicants analysis Therefore, the staff determined that the applicants selected general corrosion rate and total corrosion allowance for the internal VCC carbon steel surfaces is acceptable.

The staff reviewed the applicants evaluation of the effects of general corrosion on each of the structural, thermal, and shielding analyses described above and verified that the applicant appropriately accounted for the potential effects of reduced VCC internal steel component section thicknesses on the structural, thermal, and shielding performance of the VCC. The staff reviewed the methodology, inputs, assumptions, and conclusions of this TLAA and finds them reasonable to assess the impact of general corrosion on the VCC internal steel components safety functions; therefore, the staff finds the applicants conclusions acceptable.

The staff also notes that, regardless of the TLAA conclusion, the applicant is proposing to perform periodic inspections of the VCC under the internal and external VCC metallic monitoring AMPs to verify the condition of the steel components. These inspections are expected to inform general licensees of any unanticipated corrosion rates that may fall outside the assumptions of the TLAA, such that, using the tollgate process (discussed in SER section 3.5.1), aging management activities can be adjusted in the future, as appropriate.

Pitting and Crevice Corrosion The applicant did not provide a detailed evaluation of the effects of localized (pitting and crevice) corrosion but rather stated that the above analysis for general corrosion is applicable to localized corrosion, provided that the depth of localized corrosion does not exceed that of general corrosion. The applicant did not estimate the rate of localized corrosion, and thus the staff could not verify the bounding nature of the general corrosion analysis.

Given the lack of information on localized corrosion rates, the staffs review of the adequacy of the proposed aging management activities for localized corrosion of the steel VCC components was based on the applicants proposed AMP inspections, rather than the TLAA. The applicants AMR results cited the use of both the TLAA and the internal and external VCC metallic monitoring AMPs to manage pitting and crevice corrosion for internal and external VCC metallic components. The staff notes that NUREG-2214 recommends only an AMP to manage these aging mechanisms. As documented in SER tables 3.5-1 and 3.5-3, the staff verified that the applicants internal and external VCC metallic components monitoring AMPs are consistent with the guidance in NUREG-2214 for managing pitting and crevice corrosion of steel SCCs. On this basis, the staff determined that there is no need for the applicant to evaluate pitting and crevice corrosion as a TLAA since the applicant has included acceptable VCC metallic components monitoring AMPs for managing these aging effects, and the use of these AMPs for managing localized corrosion of VCC metallic components is consistent with the recommendations in NUREG-2214. Therefore, the staff finds that the applicants proposed aging management approach, based on implementation of the VCC metallic components monitoring AMPs, is acceptable.

3.4.3 Aging Analysis for NAC-MPC Neutron Absorber and Neutron Shield Components 3-25

3.4.3.1 Boron Depletion In section 3.3.3.3 of the renewal application, the applicant summarized its evaluation of the impact of the depletion of the boron (B)-10 content in the Boral neutron absorbers in the fuel basket and the NS-4-FR polymeric shielding material in the transfer cask and VCC shield plug on the criticality and radiation shielding safety functions of these components. Appendix B to the renewal application provides the evaluation details in proprietary Calculation No. 30013-5001, Aging Analysis for NAC-MPC/UMS Neutron Absorber and Neutron Shield Components (Storage/Transfer). The applicant provided a single analysis for both NAC-MPC and NAC-UMS systems, and it considered the most bounding case in the TLAA.

Radiation Safety The staff reviewed Calculation No. 30013-5001 as it concerns radiation shielding safety. For its baseline analysis, the applicant evaluated the NAC-UMS system spent fuel contents, since it has the highest burnup of the spent fuel contents for the NAC-MPC and NAC-UMS systems.

The burnup of the spent fuel contents for the NAC-UMS system is 60 gigawatt days per metric ton of uranium (GWd/MTU), at the same minimum enrichment as the NAC-MPC contents. The staff finds this acceptable, since the neutron source term increases with burnup, and this will maximize neutron flux and component depletion. The applicant used the XSDRN code in the SCALE 4.3 code suite to determine the neutron fluence in the shield regions. This methodology has not changed from the initial 20-year licensing-basis applications, which the staff found acceptable in its previous SERs for the NAC-MPC system. The applicants design-basis analysis determined the location of highest neutron fluence to be at the transfer cask bottom.

The applicant determined this bounding location to have a fluence 2.4 times that of the radial shield. As a conservative assumption, the applicant multiplied the radial shield depletion by a factor of 3. The staff finds this approach acceptable, as it will overpredict radial shield depletion.

The applicants design-basis neutron source was fuel burned to 40 GWd/MTU with 3.7 weight percent (wt%) initial enrichment and cooled for 5 years. The applicant scaled the neutron fluence to represent the increase in source strength for fuel at 3.7 wt% initial enrichment burned to 60 GWd/MTU and cooled for 12 years. The applicants source calculation as a function of burnup is not changed from the design-basis evaluation. The staff finds the use of a 12-year cooling time acceptable since the required cooling time for design-basis fuel is longer due to thermal constraints, and this assumption will overpredict the neutron source. The applicant determined the B-10 depletion to be less than 1 percent over the extended storage period. The staff finds reasonable assurance that B-10 depletion in the neutron shield material will be inconsequential over the period of extended storage.

Criticality Safety The staff reviewed Calculation No. 30013-5001 as it concerns criticality safety. For its baseline analysis, the applicant evaluated the NAC-UMS system spent fuel contents, since it has the highest burnup of the spent fuel contents for the NAC-MPC and NAC-UMS systems. The burnup of the spent fuel contents for NAC-UMS system is 60 GWd/MTU, at the same minimum enrichment as the NAC-MPC contents. The staff finds this acceptable, since the neutron source term increases with burnup, and this will maximize neutron fluence and component depletion.

The applicant also assumed all neutrons emitted by the design-basis fuel are absorbed in the neutron absorber sheets. This is conservative, since neutrons will escape the cask cavity and get absorbed in the neutron shield or contribute to external dose. This is also conservative, since B-10 is a thermal absorber and dry spent fuel storage is a fast neutron system. The staff finds these assumptions are acceptable since they will overpredict the B-10 depletion rate in the 3-26

neutron absorbers. The applicant determined this would result in less than 1 percent depletion of B-10 in the absorber plates after 60 years. Considering the margin provided by assuming the entire neutron fluence is absorbed by B-10, the staff finds reasonable assurance that the neutron absorber will ensure criticality safety over the period of extended storage operation.

3.4.3.2 Radiation Damage and Thermal Aging In section 3.3.3.3 of the renewal application, the applicant evaluated the potential for radiation damage and thermal aging mechanisms to affect the shielding performance of the NS-4-FR and NS-3 radiation shielding materials. For the polymeric NS-4-FR material, the applicants AMR concluded that both radiation damage and thermal aging should be evaluated as TLAAs. For the cementitious NS-3 material, the applicants AMR concluded that only radiation damage required evaluation as a TLAA. The staff noted that NUREG-2214 identified that the NS-3 material is not subject to the radiation damage and thermal aging mechanisms; however, the applicant provided a radiation damage analysis to support that conclusion for the NS-3 material.

The applicant used the results of historical radiation and thermal exposure testing of the shielding materials to define acceptable levels of exposure that do not result in material degradation and hydrogen loss. The radiation testing was performed in a research reactor and, as described below, exposed the shielding materials to levels of radiation that greatly exceed the expected exposure in the NAC-MPC System over 60 years. Calculation No. 30013-5001 contains the applicants detailed evaluation.

Radiation Damage (NS-4-FR and NS-3)

For the NS-4-FR material used in the transfer cask radial shield and VCC shield plug, the applicants calculation determined that the accumulated neutron fluence over 60 years of continuous exposure is less than 10 percent of the level that were previously shown in research reactor testing to result in no degradation and a loss of hydrogen of less than 1 percent. For gamma exposure, the applicants calculation determined that the deposited gamma radiation in the NS-4-FR material used in the transfer cask radial shield and the VCC shield plug is much less than 1 percent of the gamma exposure of this material from the research reactor tests that resulted in no significant degradation.

For the NS-3 material used in the VCC shield plug, the applicant calculated that the accumulated neutron fluence and deposited gamma radiation is less than 0.1 percent of exposure limits used in the prior research reactor testing. The applicant stated that the research reactor testing at the highest radiation exposure levels resulted in off-gassing, releasing hydrogen. To address this, the applicant considered the reduced hydrogen density for NS-3 that was used as a conservative assumption for its radiation shielding models. The applicant calculated hydrogen content after 60 years of aging by conservatively applying the measured hydrogen loss at the tested gamma radiation exposure and neutron fluence and comparing this result to the reduced hydrogen density used in its radiation shielding models. The applicant determined that the minimum remaining hydrogen content resulting from the application of the measured hydrogen loss at the tested gamma radiation exposure and neutron fluence to be about 1.5 percent lower than that used in its radiation shielding models. To address this slight non-conservative discrepancy, the applicant justified a reasonable reduction in the radiation-induced hydrogen loss for the 60-year extended storage term by considering that the research reactor testing exposed the NS-3 material to significantly greater gamma and neutron radiation than is expected in the NAC-MPC System for 60 years. With these considerations, the applicant 3-27

concluded that any potential change in the shielding performance of the NS-3 material due to radiation exposure over the 60-year extended storage term is insignificant.

The staff reviewed the applicants evaluations of the effects of radiation on the NS-4-FR and NS-3 shielding materials and verified that the estimated radiation exposure levels in the NAC-MPC System over 60 years of service are significantly lower than levels evaluated in the research reactor (less than 10 percent for NS-4-FR and less than 0.1 percent for NS-3). The staff also verified that the observed material degradation in the research reactor testing was either negligible or sufficiently limited, such that there is assurance that the shielding performance of the NS-4-FR and NS-3 materials in the NAC-MPC transfer cask radial shield and VCC shield plug will be adequately maintained for the applicants calculated 60-year radiation exposure levels. Therefore, the staff finds that the applicants TLAA adequately assesses the impact of radiation-induced aging of the neutron shield components and is, therefore, acceptable.

Thermal Aging (NS-4-FR)

For the thermal exposure of the NS-4-FR shielding material, the applicant cited the results of laboratory thermal testing at 170°C (338°F) to conclude that no significant reduction in shielding effectiveness is expected. The staff notes that the cited testing subjected the NS-4-FR material to elevated temperatures, after which the appearance, density, and gas released were characterized. No cracking or deformation was found, and loss of hydrogen was less than 1 percent. In its review of section 4 of the NAC-MPC System FSAR, the staff noted that the maximum normal exposure temperature of the NS-4-FR material in the transfer cask and VCC shield plug is significantly lower than the laboratory test exposures cited by the applicant.

Therefore, based on the results of the laboratory testing that was performed at temperatures exceeding those expected in the NAC-MPC System, the staff finds that the applicants TLAA adequately assesses the impact of thermally induced aging of the neutron shield components and is, therefore, acceptable.

3.4.4 Evaluation of Stainless-Steel Clad Spent Fuel for Fatigue in Storage In section 3.3.3.4 of the renewal application, the applicant summarized its evaluation of fatigue of the stainless steel spent fuel cladding. Appendix B to the renewal application provides the evaluation details in proprietary Calculation No. 30013-2004, Evaluation of Stainless Steel Clad Spent Fuel for Fatigue in Storage.

The applicant stated that cyclic stresses in the cladding are only expected to arise due to daily and seasonal thermal fluctuations in the cask, but it also considered the single stress cycle associated with the long-term cooling of the cladding to ambient temperature. Calculation No. 30013-2004 evaluated the stresses in the cladding using the same temperature fluctuation assumptions used in the NUREG-2214 evaluation of fatigue for zirconium-based cladding (daily variations of 25°C (77°F) and a seasonal variation of 143°C (290°F)). Two sources of stress were evaluated (internal gas pressure and cladding thermal expansion under external constraint), following the fatigue evaluation methodology in ASME B&PV Code,Section III, Division 1, NB-3222.4. For all of the temperature fluctuations (daily, seasonal, long term), the applicant found that the applied cycles were insufficient to lead to fatigue failure, based on a comparison to the relevant stainless steel fatigue curve in ASME B&PV Code,Section III, Appendix I. The applicants cumulative damage assessments also found large margins against fatigue failure.

3-28

The staff reviewed the applicants fatigue evaluation methodology and underlying assumptions (sources and severity of stressors) and verified that they are consistent with the approach recommended in NUREG-2214 and with the evaluation methodology of the ASME B&PV Code,Section III. Therefore, the staff finds that the applicant has demonstrated that fatigue will not challenge the function of the stainless steel spent fuel cladding during the period of extended operation.

3.4.5 Evaluation Findings The staff reviewed the TLAAs provided in the renewal application, following the guidance provided in NUREG-1927 and NUREG-2214. The staff verified that the TLAA inputs, assumptions, calculations, and analyses were adequate and bound the environments and aging mechanisms or aging effects for the applicable SSCs. Based on its review of the renewal application, the staff finds the following:

F3.3 The applicant identified all applicable aging mechanisms and effects for SSCs within the scope of renewal that involve TLAAs. The calculational methods and the values of the input parameters for the applicants TLAAs are adequate. Therefore, the applicants TLAAs provide reasonable assurance that the SSCs will maintain their intended functions for the period of extended operation, require no further aging management activities, and meet the requirements in 10 CFR 72.240(c)(2).

3.5 Aging Management Programs In accordance with 10 CFR 72.240(c)(3), the applicant must describe AMPs to manage issues associated with aging that could adversely affect SSCs ITS. In section 3.4 of the renewal application, the applicant proposed the following AMPs:

(1) localized corrosion and SCC of welded stainless-steel TSCs (2) internal VCCmetallic components monitoring (3) external VCCmetallic components monitoring (4) NAC reinforced VCC structuresconcrete monitoring (5) TFR and transfer adapters The staff conducted the safety review of the proposed AMPs in the renewal application according to the guidance in NUREG-1927. The staff also evaluated the proposed AMPs and compared them to the generically acceptable example AMPs in NUREG-2214, as applicable.

Tables 3.5-1 through 3.5-5 provide the staffs conclusions regarding consistency of the proposed AMPs with the applicable example AMPs in NUREG-2214. If the staff identified inconsistencies, the staffs review discusses the applicants justification.

The staff compared the applicants AMP for localized corrosion and SCC of welded stainless-steel TSCs to the example Localized Corrosion and Stress corrosion Cracking of Welded Stainless Steel Dry Storage Canisters AMP in NUREG-2214 (see table 3.5-1).

The staff compared the applicants AMPs on internal and external VCC metallic components monitoring to the example Monitoring of Metallic Surfaces AMP in NUREG-2214 (see tables 3.5-2 and 3.5-3).

3-29

The staff compared the applicants AMP for reinforced VCC structuresconcrete monitoring to the example Reinforced Concrete Structures AMP in NUREG-2214 (see table 3.5-4).

The staff compared the applicants AMP for TFRs and transfer adapters to the example Transfer Casks AMP in NUREG-2214 (see table 3.5-5).

Table 3.5-1 AMP Review ResultsLocalized Corrosion and SCC of Welded Stainless-Steel TSCs Staffs Assessment of Consistency with the Example Localized Corrosion and Stress Corrosion Cracking of AMP Element Welded Stainless Steel Dry Storage Canisters AMP in NUREG-2214

1. Scope of Program Consistent1
2. Preventive Actions Consistentthe AMP does not include preventive actions; it is a condition monitoring program.
3. Parameters Monitored or Consistent Inspected
4. Detection of Aging Effects Consistent The applicant defined inspection frequencies of every 10 years for TSCs without detection of indications of major corrosion degradation or every 5 years for TSCs with detection of major indications of corrosion degradation or detection of cracking. The guidance in NUREG-2214 does not specify an inspection frequency; however, the staff notes that the applicants proposal is consistent with ASME Code Case N-860, Inspection Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems. (2020)

ASME N-860 recommends an initial 10-year inspection interval that is reduced to 5 years in the event of a finding of major corrosion or cracking. The staff finds the applicants proposal to be acceptable, based on the lack of operating experience of canister cracking to date, consistency with ASME N-860, and the AMPs automatic adjustment to reduce inspection intervals in the event of a major finding.

5. Monitoring and Trending Consistent
6. Acceptance Criteria Consistent The proposed acceptance criteria are consistent with both the general criteria recommended in NUREG-2214 and the more detailed criteria established in ASME Code Case N-860.
7. Corrective Actions Consistent
8. Confirmation Process Consistent
9. Administrative Controls Consistent 3-30

Staffs Assessment of Consistency with the Example Localized Corrosion and Stress Corrosion Cracking of AMP Element Welded Stainless Steel Dry Storage Canisters AMP in NUREG-2214

10. Operating Experience Consistent The applicant discussed the results from two prior examinations of TSCs:
  • In 2016, a TSC containing greater than Class C waste was inspected at Maine Yankee. The inspection did not identify corrosion or cracking of the TSC. The inspection identified a small grouping of embedded iron of no appreciable depth or height. The areas were determined to be the result of iron contamination during original manufacturing or handling of the canister and, therefore, did not compromise the intended functions of the TSC.
  • In 2018, a TSC containing spent fuel was selected at Maine Yankee for meeting high susceptibility criteria and inspected in accordance with the requirements of this AMP. It was considered bounding to all TSCs in service. The inspection of the selected TSC did not have any reportable corrosion or cracking.

The applicant stated that, during the period of extended operation, each licensee will perform tollgate assessments of aggregated operating experience and other information related to the aging effects and mechanisms addressed by this AMP to determine whether changes to the AMP are required to address the current state of knowledge.

The staff reviewed the operating experience and finds that it supports the capability of the proposed AMP activities to manage aging.

1 In the evaluation for AMP element consistency with NUREG-2214, the staff considered the entirety of the applicants AMPs (i.e., in some cases, details may be captured under a different AMP element than that cited in NUREG-2214. In this instance, the AMP is nevertheless considered to be consistent with NUREG-2214).

Table 3.5-2 AMP Review ResultsInternal VCCMetallic Components Monitoring AMP Element Staffs Assessment of Consistency with the Example Monitoring of Metallic Surfaces AMP in NUREG-2214

1. Scope of Program Consistent
2. Preventive Actions Consistentthe AMP does not include preventive actions; it is a condition monitoring program.
3. Parameters Monitored or Consistent Inspected
4. Detection of Aging Effects Consistent
5. Monitoring and Trending Consistent 3-31

AMP Element Staffs Assessment of Consistency with the Example Monitoring of Metallic Surfaces AMP in NUREG-2214

6. Acceptance Criteria Consistent
7. Corrective Actions Consistent
8. Confirmation Process Consistent
9. Administrative Controls Consistent.
10. Operating Experience Consistent The applicant discussed the results from two prior examinations of internal VCC components:
  • In 2016, the internal metallic components of a VCC containing a GTCC waste canister were inspected at Maine Yankee. The inspection identified localized areas of coating damage (peeling, blistering) on the internal VCC metallic surfaces. The base metal also appeared to have minimal surface corrosion. These conditions were determined to not compromise the intended functions of the VCC.
  • In 2018, the internal metallic components of a VCC containing spent fuel were inspected at Maine Yankee. The VCC accessible internal surfaces were inspected for localized corrosion and pitting. It was estimated that 95 percent of VCC accessible surfaces were inspected. Coating deterioration and localized corrosion were identified on the liner vertical surface. The applicant reviewed the indications and evaluated potential subsequent corrosion and concluded that it did not compromise the intended functions of the VCC over a 60-year period of extended operation.

The applicant stated that, during the period of extended operation, each licensee will perform tollgate assessments of aggregated operating experience and other information related to the aging effects and mechanisms addressed by this AMP to determine whether changes to the AMP are required to address the current state of knowledge.

The staff reviewed the operating experience and finds that it supports the capability of the proposed AMP activities to manage aging.

Table 3.5-3 AMP Review ResultsExternal VCCMetallic Components Monitoring Staffs Assessment of Consistency with the Example AMP Element Monitoring of Metallic Surfaces AMP in NUREG-2214

1. Scope of Program Consistent 3-32

Staffs Assessment of Consistency with the Example AMP Element Monitoring of Metallic Surfaces AMP in NUREG-2214

2. Preventive Actions Consistentthe AMP does not include preventive actions; it is a condition monitoring program.
3. Parameters Monitored or Consistent Inspected
4. Detection of Aging Effects Consistent
5. Monitoring and Trending Consistent
6. Acceptance Criteria Consistent In addition to consistency with NUREG-2214, the AMP includes acceptance criteria that the inspection results be consistent with the calculated potential corrosion loss in the corrosion TLAA (see SER section 3.4.2).
7. Corrective Actions Consistent
8. Confirmation Process Consistent
9. Administrative Controls Consistent
10. Operating Experience Consistent The applicant stated that thousands of external VCC inspections have occurred to date on both NAC-MPC and NAC-UMS Systems (the latter being a similar system to the NAC-MPC System) as part of the past required annual inspection provision of the applicable FSAR licensing bases. The applicant summarized the findings of those results in the renewal application by citing the following:
  • No obvious metal loss has occurred to date on any VCC system.
  • Coating damage has been observed in many instances and is usually repaired in the field as part of a coating touchup campaign. The licensee schedules this activity at convenient intervals and during optimum weather conditions. At no time has coating damage led to obvious metal loss.
  • In 2018, the external metallic components of a VCC were inspected at Maine Yankee. The inspection of the selected VCC did not identify any significant corrosion or loss of base metal.

The applicant stated that, during the period of extended operation, each licensee will perform tollgate assessments of aggregated operating experience and other information related to the aging effects and mechanisms addressed by this AMP to determine whether changes to the AMP are required to address the current state of knowledge.

The staff reviewed the operating experience and finds that it supports the capability of the proposed AMP activities to manage aging.

3-33

Table 3.5-4 AMP Review ResultsVCC StructuresConcrete Monitoring Staffs Assessment of Consistency with the Example AMP Element Reinforced Concrete Structures AMP in NUREG-2214

1. Scope of Program Consistent
2. Preventive Actions Consistent Although not cited in the applicants AMP, the NAC-MPC System technical specifications require actions to ensure that the heat removal capabilities of the VCC are maintained (e.g., verifying that vents are unobstructed). This is consistent with a preventive action cited in NUREG-2214 to ensure that concrete is not exposed to temperatures that could cause degradation.
3. Parameters Monitored or Consistent Inspected
4. Detection of Aging Effects Consistent
5. Monitoring and Trending Consistent
6. Acceptance Criteria Consistent
7. Corrective Actions Consistent
8. Confirmation Process Consistent
9. Administrative Controls Consistent
10. Operating Experience Consistent The applicant stated that thousands of VCC inspections have occurred to date on both NAC-MPC and NAC-UMS Systems (the latter being a similar system to the NAC-MPC System) as part of the past required annual inspection provision of the applicable FSAR licensing bases. The applicant summarized the findings of those results in the renewal application by citing the following:
  • Passive cracking has been observed, which has been attributed to shrinkage cracking during construction. The cracks that have been trended have not changed in size, shape, or extent.
  • Spalling has been observed at cold weather sites, which has been attributed to the forces associated with thermal expansion differences between the concrete and the base plate or the prying action of freeze thaw damage. It is an active mechanism for spalling.
  • Efflorescence has been observed to varying degrees at different sites. It is generally considered benign and has not been associated with concrete degradation.
  • No staining or spalling due to rebar corrosion has been identified in the loaded fleet.

3-34

The applicant stated that, during the period of extended operation, each licensee will perform tollgate assessments of aggregated operating experience and other information related to the aging effects and mechanisms addressed by this AMP to determine whether changes to the AMP are required to address the current state of knowledge.

The staff reviewed the operating experience and finds that it supports the capability of the proposed AMP activities to manage aging.

Table 3.5-5 AMP Review ResultsTFRs and Transfer Adapters Staffs Assessment of Consistency with the Example AMP Element Transfer Casks AMP in NUREG-2214

1. Scope of Program Consistent The applicant explained that the AMP is not applicable to facilities not maintaining a TFR or transfer adapter on site.
2. Preventive Actions Consistent
3. Parameters Monitored or Consistent Inspected
4. Detection of Aging Effects Consistent
5. Monitoring and Trending Consistent
6. Acceptance Criteria Consistent
7. Corrective Actions Consistent
8. Confirmation Process Consistent
9. Administrative Controls Consistent
10. Operating Experience Consistent. The applicant stated that, during the period of extended operation, each licensee will perform tollgate assessments of aggregated operating experience and other information related to the aging effects and mechanisms addressed by this AMP to determine whether changes to the AMP are required to address the current state of knowledge.

The applicant stated that, during the periods of use of the TFRs and transfer adapters at the licensees facilities, the TFRs were maintained and inspected in accordance with the requirements of American National Standards Institute (ANSI) N14.6, American National Standard for Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More (ANSI, 1993). During operation of the TFRs and transfer adapters, areas of coating degradation were repaired by reapplication of coatings. No issues with general, pitting, crevice, or galvanic corrosion have been identified. No excessive wear or loss of material 3-35

has been identified on shield door-to-door rail-to-transfer adapter surfaces. No cracking of TFR lifting trunnions has been identified.

The staff reviewed the operating experience and finds that it supports the capability of the proposed AMP activities to manage aging.

3.5.1 Aging Management Tollgates The applicant incorporated periodic tollgate assessments as requirements in the renewed CoC as recommended in NEI 14-03, revision 2. The schedule for these tollgate assessments will be incorporated into chapter 14 of the UFSAR.

The staff noted that the purpose of the tollgate concept is to provide a structured way for licensees to formally assess aggregated aging management feedback at specific points in time during the period of extended storage and perform a safety assessment that confirms the safe storage of spent nuclear fuel.

The applicant stated that general licensees are required to perform and document periodic tollgate assessments on the state of knowledge of aging-related operational experience, research, monitoring, and inspections to ascertain the ability of in-scope NAC-MPC System SSCs to continue to perform their intended safety functions throughout the renewed period of extended operation. The applicant explained that the general requirements for the periodic tollgate assessments must be addressed in the programs and procedures that are established, maintained, and implemented by each general licensee for the AMPs.

The applicant stated that each general licensee shall complete the initial tollgate assessment within 5 years following the 20th inservice year of the first NAC-MPC System loaded at each site or 6 years after the effective date of the CoC renewal, whichever is later. Subsequent tollgate assessments will be performed at a 10-year (+/- 1 year) frequency thereafter. The initial tollgate assessment is timed to allow the initial round of AMP inspections to be completed at the general-licensed site before the initial tollgate assessment, such that the operating experience gained from the initial round of AMP inspections can be evaluated and assessed. The applicant further stated that the 10-year frequency for subsequent tollgate assessments reflects the risk significance of the aging effects managed by AMPs. The applicant explained that, if the results of previous tollgate assessments indicate unanticipated or accelerated aging effects, the period for follow-up assessments will be reduced based upon the timing of the aging mechanisms identified and their risk significance. The basis of any adjustments in the tollgate assessment frequency shall be included in the tollgate assessment report.

The applicant stated that, at a minimum, the periodic tollgate assessments to be performed by each general licensee shall consider the operating experience related to the aging effects managed by the AMPs from the general licensees completed inspections and those of other general licensees that use the NAC-MPC System. The assessments will also consider new information on relevant aging effects from related industry operating experience, research findings, monitoring data and inspection results, NRC generic communications, U.S. Department of Energy research updates, Aging Management Institute of Nuclear Power Operations (INPO) Database (AMID), and relevant information and reports from industry organizations such as the NEI, EPRI, and INPO, as applicable. The aggregated operating experience will be evaluated to identify any new aging effects or aging mechanisms that may be applicable to the in-scope SSCs of the NAC-MPC System or are not adequately managed by 3-36

the current AMPs or TLAAs. The assessment will also evaluate if continued safe storage is expected until the next tollgate assessment, or if additional aging management activities are required to address newly identified aging effects requiring management.

The applicant further stated that each general licensee shall document the periodic tollgate assessment in a report, which will include the following information, at a minimum:

  • the sources of operating experience, aggregated research findings, monitoring data, and inspection results considered in the assessment
  • a summary of the research findings, operating experience, monitoring data and inspection results
  • the potential impact, if any, of the research findings, operating experience, monitoring data, and inspection results on the AMPs or TLAAs for the in-scope SSCs
  • recommended corrective actions to be implemented to address newly identified aging effects that are not adequately managed by the existing AMPs or TLAAs
  • summary and conclusions The applicant explained that the general licensee will maintain the tollgate assessment report(s) as a permanent record, in accordance with the requirements of its quality assurance program and will be available for NRC inspection. A copy of each tollgate assessment report will also be provided to the CoC holder. As deemed appropriate, an industry organization (e.g., NEI, EPRI, INPO) will disseminate the tollgate assessment reports.

The staff reviewed the applicants description of actions to ensure that the AMP remains adequate during the period of extended operation upon review of new operating experience.

The staff considers that the implementation of periodic tollgate assessments and the use of the AMID database, in addition to other periodic operating experience reviews consistent with the site quality assurance program, provide reasonable assurance that the applicants AMPs will remain adequate during the period of extended operation.

3.5.2 Evaluation Findings The staff reviewed the AMPs provided in the renewal application. The staff performed its review following the guidance in NUREG-1927 and NUREG-2214. Based on its review, the staff finds the following:

F3.4 The applicant has identified programs that provide reasonable assurance that aging mechanisms and effects will be managed effectively during the period of extended operation, in accordance with 10 CFR 72.240(c)(3).

3-37

4 CHANGES TO CERTIFICATE OF COMPLIANCE AND TECHNICAL SPECIFICATIONS This section provides a consolidated list of the changes to the CoC conditions and technical specifications resulting from the review of the renewal application. The basis of the changes is provided here for those changes that are not described elsewhere in this SER.

Changes to the Certificate of Compliance

1. Added the following condition to the initial CoC (Amendment 0) and Amendments 1-8:

FSAR UPDATE FOR RENEWED CoC The CoC holder shall submit an updated final safety analysis report (FSAR) to the Commission, in accordance with 10 CFR 72.4, within 90 days of the effective date of the CoC renewal. The updated FSAR shall reflect the changes resulting from the review and approval of the CoC renewal. The CoC holder shall continue to update the FSAR pursuant to the requirements of 10 CFR 72.248.

The CoC holder has indicated that changes will be made to the updated FSAR to address aging management activities resulting from the renewal of the CoC. This condition ensures that the updated FSAR changes are made in a timely fashion to enable general licensees using the storage system during the period of extended operation to develop and implement necessary procedures.

The CoC holder proposed changes in FSAR Section 8.0, Operating Procedures that outlined administrative controls intended to clarify the approach to various short-term operations (ML22203A127). After additional consideration, in a subsequent supplement (ML22355A119) the CoC holder requested that those administrative controls no longer be included in the renewal application. Staff confirms that these administrative controls are not considered to be included in this CoC condition.

2. Added the following condition to the initial CoC (Amendment 0) and Amendments 1-8:

10 CFR 72.212 EVALUATIONS FOR RENEWED CoC USE Any general licensee that initiates spent fuel dry storage operations with the NAC-MPC System after the effective date of the CoC renewal and any general licensee operating a NAC-MPC 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.

4-1

b. As part of the document review required by 10 CFR 72.212(b)(6),

include a review of the FSAR changes resulting from the renewal of the CoC and the NRC Safety Evaluation Report related to the renewal of the CoC.

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 this Condition 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 staff considers it important to ensure that appropriate considerations for the period of extended operation are evaluated in the general licensees report required by 10 CFR 72.212, Conditions of general license issued under § 72.210. These considerations arise from the analyses and assumptions in the renewal application regarding operations during the period of extended operation. This includes potential use by general licensees that may use a new NAC-MPC System after the CoC has been renewed either at a new or an existing general-licensed ISFSI. The renewal of the CoC is based on assumptions and analyses of the dry storage system and the sites where it is used. Licensees considering the use of the NAC-MPC System must evaluate it for use at their respective sites. This condition also makes it clear that general licensees that currently use an NAC-MPC System will need to update their 10 CFR 72.212 reports, even if they do not put additional dry storage systems into service after the renewals effective date, in accordance with 10 CFR 72.212(b)(11).

3. Added the following condition to the initial CoC (Amendment No. 0) and Amendments 1-8:

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 they remain adequate for any changes to SSCs within the scope of the CoC renewal.

The applicant proposed the above CoC condition in appendix D to the renewal application. The staff recognizes that the CoC may continue to be amended after it has been renewed. This condition ensures that future amendments to the CoC address the renewed design bases for the CoC, including aging management considerations that may arise from the changes to the system in proposed future amendments.

4. Revised initial CoC (Amendment 0) and Amendments Nos. 1-8 to address change to language in 10 CFR 72.210 and other updates to the regulations:

This change is made for consistency with the language currently in 10 CFR 72.210 and other cited regulations and is not pertinent to the safety review conducted for the renewal application.

Changes to the text are in bold, which only involves adding new text.

4-2

AUTHORIZATION The NAC-MPC System, which is authorized by this certificate, is hereby approved for general use by holders of 10 CFR Part 50 and 10 CFR Part 52 licenses for nuclear reactors at reactor sites under the general license issued pursuant to 10 CFR 72.210, subject to the conditions specified by 10 CFR 72.212, and the attached appendix A and appendix B.

Any CoC and technical specification language that discusses licensees under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, was modified to also include licensees under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. In addition, updates to the regulation citations referenced in the applicable CoC and technical specifications have been changed to reflect citations currently in the regulations.

CHANGES TO TECHNICAL SPECIFICATIONS

1. Added section A.5.X to appendix A to the technical specifications associated with the initial CoC (Amendment No. 0) and Amendments Nos. 1-8:

Added the new technical specification [A.5.X]:

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 (FSAR). 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 procedure, and that reference shall be maintained even if the procedure 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 NAC-MPC Systems in service for longer than 20 years.

The CoC holder proposed a condition to revise or create programs or procedures for implementing the AMPs in the supplement to the FSAR. This specification ensures that programs or procedures address AMP activities required for extended storage operations.

4-3

5 CONCLUSION Pursuant to 10 CFR 72.240(d), the NRC will renew the CoC of a design of a spent fuel storage cask if (1) the quality assurance requirements in 10 CFR Part 72, Subpart G, Quality Assurance, are met, (2) the requirements of 10 CFR 72.236(a) through (i) are met, and (3) the application includes a demonstration that the storage of spent fuel has not, in a significant manner, adversely affected ITS SSCs. Additionally, 10 CFR 72.240(c) requires that the safety analysis report accompanying the application contain TLAAs and AMPs that demonstrate that the dry storage system SSCs will continue to perform their intended functions for the requested period of extended operation.

The NRC staff reviewed the renewal application for the NAC-MPC System, in accordance with NRC regulations in 10 CFR Part 72. The staff followed the guidance in NUREG-1927, Revision 1. Based on its review of the renewal application and the CoC conditions, the staff determines that the dry storage system has met the requirements of 10 CFR 72.240.

5-1

6 REFERENCES American National Standards Institute, ANSI N14.6, American National Standard for Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More, New York, New York, 1993.

American Society of Mechanical Engineers (ASME) Code Case N-860, Inspection Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transport Containment Systems, New York, New York, 2020.

Electric Power Research Institute (EPRI), Technical Report 106440, Evaluation of Expected Behavior of LWR Stainless Steel-Clad Fuel in Long-Term Dry Storage, prepared by Pacific Northwest National Laboratories, Richland, Washington, 1996.

NAC International, Submission of a Request to Renew the U.S. Nuclear Regulatory Commission Certificate of Compliance No. 1025 for the NAC-MPC Cask System, Norcross, Georgia, December 2019. Agencywide Documents Access and Management System (ADAMS)

Accession No. ML19357A178.

_____. NAC-MPC Biennial FSAR Update and Associated 10 CFR 72.48 Determination Summary Report, Norcross, Georgia, April 2020. ML20108E871.

_____. Submittal of Responses to the Nuclear Regulatory Commissions (NRC) Request for Additional Information for the Request to Renew the NRC Certificate of Compliance No. 1025 for the NAC-MPC Cask System, Norcross, Georgia, August 2021. ML21231A154.

_____. Submission of Responses to the Nuclear Regulatory Commissions (NRC) Request for Additional Information for the Request to Renew the NRC Certificate of Compliance No. 1025 for the NAC-MPC Cask System, Norcross, Georgia, March 2022. ML22077A832.

_____. Supplement to the Submission of Responses to the Nuclear Regulatory Commissions (NRC) Request for Additional Information for the Request to Renew the NRC Certificate of Compliance No. 1025 for the NAC-MPC Cask System, Norcross, Georgia, July 2022.

ML22203A127.

_____. Supplement to the Submission of Responses to the Nuclear Regulatory Commissions (NRC) Request for Additional Information for the Request to Renew the NRC Certificate of Compliance No. 1025 for the NAC-MPC Cask System, Norcross, Georgia, December 2022.

ML22355A119.

National Association of Corrosion Engineers. Corrosion Engineers Reference Book, 4th Edition, Houston, Texas, 2016 Nuclear Energy Institute (NEI), NEI 14-03, Revision 2, Guidance for Operations-Based Aging Management for Dry Cask Storage, Washington, DC, 2016.

_____. NEI-22-02, Guidelines for Weather-Related Administrative Controls for Short Duration Outdoor Dry Cask Storage Operations, Washington, DC, 2022.

6-1

U.S. Nuclear Regulatory Commission (NRC), NUREG-1927, Revision 1, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Washington, DC, June 2016. ML16179A148.

_____. NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, Final Report, Washington, DC, July 2019. ML19214A111.

_____. Safety Evaluation Report, NAC Multi-Purpose Canister (NAC-MPC) System, Washington, DC, March 2000. ML003704104.

_____. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Washington, DC.

_____. CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, Washington, DC.

_____. 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, Washington, DC.

Olender, A., J. Gorman, C. Marks, and G. Ilevbare, Recent Operating Experience Issues with 17-4 PH in LWRs, Fontevraud 8: Conference on Contribution of Materials Investigations and Operating Experience to LWRs Safety, Performance and Reliability, France, 2015.

6-2

Retrieved from "https://kanterella.com/w/index.php?title=ML23255A254&oldid=3618189"