RA-20-0317, Review Request for the Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227, Revision 1-A

Review Request for the Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227, Revision 1-A

ML20294A110
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Site:	Catawba
Issue date:	10/15/2020
From:	Simril T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML20294A109	List: RA-20-0317, Review Request for the Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227, Revision 1-A
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RA-20-0317 WCAP-18496-NP, Rev. 0
Download: ML20294A110 (127)

Text

Tom Simril

( ~ DUKE Vice President ENERGY Catawba Nuclear Station Duke Energy CN01VP I 4800 Concord Road York, SC 29745 o: 803.701.3340 f: 803.701.3221 RA-20-0317 October 15, 2020 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Review Request for the Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227, Revision 1-A Catawba's Updated Final Safety Analysis Report (UFSAR) Section 18.2.22 contains the existing reactor vessel internals (RVI) inspection commitments. This UFSAR section contains an allowance that permits Duke Energy to modify or eliminate these inspections if plant-specific justification is provided to demonstrate the basis for the modification or elimination. Additionally, as part of its license renewal, Duke Energy stated they would participate in industry activities associated with RVl-related issues and that the Catawba RVI Program is subject to future enhancements as the industry's understanding of degradation continues to improve. The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI. The results of the industry recommended inspections serve as the basis for identifying any augmented inspections that are required to complete the Catawba RVI Program.

By letter dated June 16, 2010, Duke Energy notified the Nuclear Regulatory Commission (NRC) of its intent to revise its commitments for RVI inspections from those that currently exist in Catawba's UFSAR to the inspection guidelines provided by Materials Reliability Program (MRP)-227 as approved by the NRC (i.e., MRP-227, Revision 1-A).

By letter dated March 19, 2014, Duke Energy notified the NRC of its intent to submit a plant specific RVI inspection plan for Catawba to implement MRP-227-A (or the latest NRC approved revision) no later than two years before the initial inspection.

Attachment 2 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of proprietary information, this letter is uncontrolled.

United States Nuclear Regulatory Commission Page 2 October 15, 2020 contains the Aging Management Program (AMP) and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals (Non-Proprietary). This AMP and Inspection Plan provides a description of the Catawba Reactor Vessel Internals Program and is based on MRP-227, Revision 1-A as a strategy for managing age-related material degradation in reactor vessel internal components through the period of extended operation. Once this document is approved by the NRC, the Catawba UFSAR will be updated as required.

Section 6 of Attachment 1 details how Catawba meets the applicability requirements contained in Section 2.4 and Appendix B of MRP-227, Revision 1-A. Appendix B of MRP-227, Revision 1-A provides guidance on plant-specific fuel design and fuel management requirements. Attachment 2 is intended to address Appendix B of MRP-227, Revision 1-A and supplement the discussion contained in Section 6 of Attachment 1. contains the Catawba Units 1 & 2 Summary Report for the Fuel Design/Fuel Management Assessment to Demonstrate MRP-227, Revision 1-A, Applicability (Proprietary and Non-Proprietary, reference the Attachments within this document). contains the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-20-5005, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Attachment 2 contains information proprietary to Westinghouse Electric Company LLC

("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the NRC's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the NRC's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-20-5005 and should be addressed to Camille T.

Zozula, Manager, Infrastructure & Facilities Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Suite 165, Cranberry Township, Pennsylvania 16066.

Catawba Nuclear Station is making a one-time regulatory NRC commitment to incorporate in the Catawba UFSAR a description of the Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to implement MRP-227, Revision 1-A). The due date for this NRC commitment is the next periodic update of the UFSAR in accordance with 10 CFR 50.71(e) after NRC approval. This is the only regulatory NRC commitment listed in this letter. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.

Attachment 2 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of proprietary information, this letter is uncontrolled.

United States Nuclear Regulatory Commission Page3 October 15, 2020 Duke Energy requests NRC approval of this submittal by October 15, 2021, to support MRP-227 implementation activities. If you have any questions or require additional information, please contact Sherry Andrews of Regulatory Affairs at (803) 701-3424.

Sincerely, Tom Simril Vice President, Catawba Nuclear Station Attachments Attachment 2 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of proprietary information, this letter is uncontrolled.

United States Nuclear Regulatory Commission Page 4 October 15, 2020 xc (with attachment):

Laura Dudes Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303 Karen Cotton NRC Project Manager (CNS)

U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop O-8G9A Rockville, MD 20852 Joseph Austin NRC Senior Resident Inspector Catawba Nuclear Station Attachment 2 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of proprietary information, this letter is uncontrolled.

ATTACHMENT 1 Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227, Revision 1-A)

WCAP-18496-NP, Rev. 0 (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-18496-NP September 2020 Revision 0 Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227, Revision 1-A)

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-18496-NP Revision 0 Aging Management Program and Inspection Plan for the Catawba Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227, Revision 1-A)

Author: Taylor N. Zindren*

Materials & Aging Management Verifier: Joshua K. McKinley*

Materials & Aging Management September 2020 Approved: Kaitlyn M. Musser*, Manager Materials & Aging Management This document may contain technical data subject to the export control laws of the United States. In the event that this document does contain such information, the Recipients acceptance of this document constitutes agreement that this information in document form (or any other medium),

including any attachments and exhibits hereto, shall not be exported, released or disclosed to foreign persons whether in the United States or abroad by recipient except in compliance with all U.S. export control regulations. Recipient shall include this notice with any reproduced or excerpted portion of this document or any document derived from, based on, incorporating, using or relying on the information contained in this document.

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2020 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii TABLE OF CONTENTS LIST OF TABLES........................................................................................................................................................ iv LIST OF FIGURES ....................................................................................................................................................... v LIST OF ACRONYMS ................................................................................................................................................ vi ACKNOWLEDGEMENTS ........................................................................................................................................viii 1 PURPOSE ....................................................................................................................................................1-1 2 BACKGROUND ..........................................................................................................................................2-1 2.1 INDUSTRY EFFORT....................................................................................................................2-1 2.2 CATAWBA REACTOR VESSEL INTERNALS .........................................................................2-2 2.3 CATAWBA LICENSE RENEWAL ..............................................................................................2-4 3 PROGRAM OWNER ..................................................................................................................................3-1 4 DESCRIPTION OF CATAWBA REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS .................................................................................................................4-1 4.1 SUPPORTING CATAWBA PROGRAMS AND ACTIVITIES ...................................................4-3 4.1.1 Chemistry Control Program ........................................................................................ 4-4 4.1.2 ASME Section XI Program ......................................................................................... 4-4 4.1.3 Bottom-Mounted Instrumentation Thimble Tube Inspection ...................................... 4-4 4.1.4 Control Rod Guide Tube Support Pin Replacement Project ....................................... 4-5 4.1.5 Power Uprating Project for Catawba Unit 1................................................................ 4-6 4.2 INDUSTRY PROGRAMS ............................................................................................................4-6 4.2.1 WCAP-14577, Aging Management for Reactor Internals .......................................... 4-6 4.2.2 MRP-227, Revision 1-A, Reactor Internals Inspection and Evaluation Guidelines .... 4-6 4.2.3 WCAP-17451-P, Reactor Internals Guide Tube Wear ................................................ 4-8 4.2.4 Baffle-Former Bolt Degradation ................................................................................. 4-9 4.2.5 WCAP-17096, Reactor Internals Acceptance Criteria Methodology and Data Requirements ............................................................................................................ 4-10 4.2.6 Core Barrel Operating Experience ............................................................................ 4-10 4.2.7 Clevis Bearing Stellite Wear Surface and Clevis Insert Bolts - Technical Bulletin TB-14-5 ..................................................................................................................... 4-11 4.2.8 Thermal Sleeve Flange Wear .................................................................................... 4-11 4.2.9 On-Going Industry Programs and NEI 03-08 Guidelines ......................................... 4-12 4.3

SUMMARY

.................................................................................................................................4-13 5 CATAWBA NUCLEAR STATION UNIT 1 AND UNIT 2 REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES ...........................................................................................5-1 5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM ......................................................5-1 5.2 GALL REVISION 2 ELEMENT 2: PREVENTIVE ACTIONS ..................................................5-3 5.3 GALL REVISION 2 ELEMENT 3: PARAMETERS MONITORED OR INSPECTED .............5-4 5.4 GALL REVISION 2 ELEMENT 4: DETECTION OF AGING EFFECTS .................................5-5 5.5 GALL REVISION 2 ELEMENT 5: MONITORING AND TRENDING ....................................5-8 5.6 GALL REVISION 2 ELEMENT 6: ACCEPTANCE CRITERIA .............................................5-10 5.7 GALL REVISION 2 ELEMENT 7: CORRECTIVE ACTIONS ................................................5-10 5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS ..........................................5-11 5.9 GALL REVISION 2 ELEMENT 9: ADMINISTRATIVE CONTROLS ...................................5-12 WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii 5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE .........................................5-13 6 MRP-227 SAFETY EVALUATION CONDITIONS AND ACTION ITEMS ...........................................6-1 6.1 MRP-227, REVISION 1-A GUIDELINE APPLICABILITY .......................................................6-1 6.1.1 General Assumptions .................................................................................................. 6-2 6.1.2 MRP-191, Revision 2 FMECA ................................................................................... 6-6 6.1.3 U.S. NRC Acceptance of MRP-227, Revision 1-A..................................................... 6-8 6.2 MRP-227, REVISION 1 SE APPLICANT/LICENSEE ACTION ITEM 1: DEGRADATION OF BAFFLE-FORMER BOLTS .................................................................................................6-10 7 INSPECTION PLAN AND IMPLEMENTATION SCHEDULE ...............................................................7-1 8

SUMMARY

AND CONCLUSIONS ...........................................................................................................8-1 9 REFERENCES .............................................................................................................................................9-1 APPENDIX A ILLUSTRATIONS........................................................................................................................ A-1 APPENDIX B CATAWBA UNIT 1 AND UNIT 2 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLES ................................................................................................................ B-1 APPENDIX C MRP-227 AUGMENTED INSPECTIONS .................................................................................. C-1 WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 6-1: Reactor Internals FMECA (Significance) Groups Used by the MRP-191 Revision 2 Expert Panel ............................................................................................................................. 6-7 Table 7-1: Catawba Unit 1 Primary Component Inspection Plan ............................................................. 7-2 Table 7-2: Catawba Unit 2 Primary Component Inspection Plan ............................................................. 7-4 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components) ........................................................................................................................... B-1 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals........................................................................................... C-1 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals........................................................................................... C-6 Table C-3: MRP-227, Revision 1-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals ................................................ C-11 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals ................................................................................... C-13 WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v LIST OF FIGURES Figure A-1: Illustration of Typical Westinghouse Internals ..................................................................... A-1 Figure A-2: Typical Westinghouse Control Rod Guide Card ................................................................... A-2 Figure A-3: Typical Lower Section of Control Rod Guide Tube Assembly ............................................ A-3 Figure A-4: Major Core Barrel Welds ...................................................................................................... A-4 Figure A-5: Bolting Systems Used in Westinghouse Core Baffles .......................................................... A-5 Figure A-6: Core Baffle/Barrel Structure ................................................................................................. A-6 Figure A-7: Bolting in a Typical Westinghouse Baffle-Former Structure................................................ A-7 Figure A-8: Lower Core Support Structure.............................................................................................. A-8 Figure A-9: Lower Core Support Structure - Core Support Plate Cross-Section .................................... A-9 Figure A-10: Typical Core Support Column .......................................................................................... A-10 Figure A-11: Typical Bottom-Mounted Instrumentation (BMI) Column Design .................................. A-11 Figure A-12: Typical Westinghouse-Design Upper Internals Assembly Upper Core Plate ................... A-12 WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi LIST OF ACRONYMS A/LAI Applicant/Licensee Action Item AMP Aging Management Program AMR Aging Management Review ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BMI bottom-mounted instrumentation CASS cast austenitic stainless steel CE Combustion Engineering CFR Code of Federal Regulations CLB current licensing basis CRDM control rod drive mechanism CRGT control rod guide tube CSB core support barrel EFPY effective full-power year EPRI Electric Power Research Institute ET eddy current testing EVT enhanced visual testing (a visual NDE method that includes EVT-1)

FMECA failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation-assisted stress corrosion cracking ID inner diameter IE irradiation embrittlement INPO Institute of Nuclear Power Operations ISI inservice inspection ISR irradiation-enhanced stress relaxation LAR License Amendment Request LR license renewal LRA license renewal application MRP Materials Reliability Program MWt megawatts thermal NDE nondestructive examination NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system OD outer diameter OE operating experience OEM original equipment manufacturer PH precipitation-hardening PWR pressurized water reactor PWROG Pressurized Water Reactor Owners Group PWSCC primary water stress corrosion cracking QA quality assurance WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vii LIST OF ACRONYMS (cont.)

RAI request for additional information RCCA rod cluster control assembly RCS reactor coolant system RVI reactor vessel internals SCC stress corrosion cracking SE Safety Evaluation SER Safety Evaluation Report SRP Standard Review Plan SS stainless steel TE thermal embrittlement UFSAR Updated Final Safety Analysis Report UHI upper head injection U.S. United States UT ultrasonic testing (a volumetric NDE method)

VT visual testing (a visual NDE method that includes VT-1 and VT-3)

WOG Westinghouse Owners Group XL Extra-long Westinghouse Fuel WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 viii ACKNOWLEDGEMENTS The authors would like to thank the members of the Duke Energy Corporation, Aging Management Program Team led by Rachel Doss and Chris Mallner and our associates at Westinghouse for their efforts in supporting the development of this WCAP.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 1 PURPOSE The purpose of this report is to document the Duke Energy, Catawba Nuclear Station Unit 1 and Catawba Nuclear Station Unit 2 (hereafter referred to as Catawba) Reactor Vessel Internals (RVI) Aging Management Program (AMP) and Inspection Plan for submittal to the United States (U.S.) Nuclear Regulatory Commission (NRC). The purpose of the Catawba RVI Program is to manage the effects of aging on RVI through the license renewal period (which begins at midnight on December 6, 2024 for Catawba Unit 1 and at midnight on February 24, 2026 for Catawba Unit 2). This document demonstrates that the Catawba RVI Program manages the effects of aging for RVI components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function throughout the Catawba period of extended operation. This document is supported by existing Duke Energy documents and procedures. As needed by industry experience or directive in the future, the Catawba RVI Program will be updated or supported by additional documents to provide clear and concise direction for the effective management of aging degradation in RVI components. These actions provide assurance that operations at Catawba will continue to be conducted in accordance with the current licensing basis (CLB) for the RVI by fulfilling license renewal [1], American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI Inservice Inspection (ISI) programs [39, 40, and 41] (and the applicable Section XI Edition and Addenda [2]), and industry requirements [42]. The Catawba RVI Program fully captures the intent of the industry guidance for RVI augmented inspections, based on the programs sponsored by U.S. utilities through the Electric Power Research Institute (EPRI) managed Materials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG).

The main objectives for the Catawba RVI AMP and Inspection Plan are to:

• Demonstrate that the effects of aging on the RVI will be adequately managed for the period of extended operation in accordance with 10 CFR 54 [4].

• Summarize the role of relevant existing Catawba programs and activities in the RVI Program.

• Define the Catawba RVI Program, based on industry-defined (EPRI/MRP and PWROG) RVI requirements and guidance, addressing the ten (10) AMP elements in NUREG-1801, Revision 2

[6] as updated by LR-ISG-2011-04 [38].

• Address the applicable Applicant/Licensee Action Item (A/LAI) identified in the Safety Evaluation (SE) on MRP-227, Revision 1-A (contained in MRP-227, Revision 1-A) [51], as well as the guideline of applicability within Section 2.4 and guidance on plant-specific applicability related to fuel design or fuel management within Appendix B of MRP-227, Revision 1-A.

• Provide an inspection plan for the Catawba reactor internals.

During the review of the License Renewal Application [1], Duke Energy made commitments related to the Catawba RVI Program. In a letter dated June 16, 2010 [32], Duke Energy notified the NRC of its intent to revise its commitments for RVI inspections from those that currently exist in Catawbas Updated Final Safety Analysis Report (UFSAR) to the inspection guidelines provided by MRP-227 as approved by the NRC [32]. The existing inspection commitments are contained in Section 18.2.22 of the Catawba WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-2 UFSAR. The UFSAR section contains an allowance that permits Duke Energy to modify or eliminate these inspections if plant-specific justification is provided to demonstrate the basis for the modification or elimination. Additionally, as part of its license renewal, Duke Energy stated that they would participate in industry activities associated with RVI-related issues and that the Catawba RVI Program is subject to future enhancements as the industry understanding of degradation continues to change. The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI. The results of the industry-recommended inspections serve as the basis for identifying augmented inspections that are required to complete the Catawba RVI Program. Duke Energy is revising its commitments for RVI Inspections from those that currently exist in the Catawba UFSAR to the inspection guidelines provided in MRP-227, as approved by the NRC [32]. Since the June 16, 2010 letter [32], MRP-227 has been revised to Revision 1 and has received NRC approval through the MRP-227 Revision 1 SER issued on April 25, 2019 [52]. Therefore, this CNS RVI AMP and Inspection Plan, as documented herein, is based on MRP-227, Revision 1-A. Once the Catawba RVI AMP and Inspection Plan is approved by the NRC, the Catawba UFSAR will be updated as required.

In a letter dated March 19, 2014 [28], Duke Energy notified the NRC of its intent to submit a plant-specific RVI inspection plan for Catawba to implement MRP-227-A (or latest NRC approved revision, currently MRP-227, Revision 1-A) no later than 2 years before the initial inspection. Within [28], Duke Energy also notified the NRC that the expected inspection plan submittal date for Catawba is Fall 2022.

The expected initial inspection dates are Fall 2024 and Fall 2025 for Catawba Units 1 and 2, respectively, but may be subject to change.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 2 BACKGROUND The management of aging degradation effects in RVI is required for nuclear plants considering or entering license renewal, as specified in NUREG-1800, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants (SRP-LR) [7]. To accomplish this task the nuclear industry has been engaged in several efforts to provide general guidelines to manage the aging of RVI for the industry as a whole. On a plant-specific basis, Catawba has defined its RVI and has demonstrated that the aging effects associated with the Catawba RVI will be adequately managed throughout the period of extended operation.

2.1 INDUSTRY EFFORT The U.S. nuclear power industry has been actively engaged in supporting the goal of managing aging degradation effects in RVI. Various programs have been underway within the industry over the past decades to develop guidelines for managing the effects of aging within pressurized water reactor (PWR) internals. In 1997, the Westinghouse Owners Group (formerly WOG, now PWROG) issued WCAP-14577, License Renewal Evaluation: Aging Management for Reactor Internals, which was reissued as Revision 1-A in 2001 after receiving NRC staff review and approval [8]. Other efforts were engaged by the EPRI MRP to address the PWR internals aging management issue for the following three currently operating domestic reactor designs: Westinghouse, Combustion Engineering (CE), and Babcock &

Wilcox (B&W).

The MRP first established a framework and strategy for the aging management of PWR internals components using proven and familiar methods for inspection, monitoring, surveillance, and communication. Based upon that framework and strategy, as well as accumulated industry research data, the following elements were further developed [94], [36], and [93]:

• Screening criteria, considering material properties, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR internals components to each of eight (8) postulated aging mechanisms (further discussed in Section 4 of this document)

• Categorization of PWR internals components, based on the screening criteria and the likelihood and severity of safety and economic consequences, into groups: components with insignificant effects from aging degradation, components with the potential to have moderately significant effects from aging degradation, and components with the potential to be significantly affected by aging degradation

• Functionality assessments to determine the effects of the degradation mechanisms on component functionality based on representative plant designs of PWR internals components and assemblies of components using irradiated and aged material properties.

Aging management strategies were developed by combining the results of the functionality assessment with several additional factors to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. The additional factors considered WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 2-2 included component accessibility, operating experience (OE), existing evaluations, and prior examination results [95].

The industry effort, as coordinated by the EPRI MRP, has finalized Inspection and Evaluation (I&E)

Guidelines for RVI. The industry guidance is contained within two (2) separate EPRI MRP documents:

• MRP-227, Revision 1-A, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines [51] (hereafter referred to as the I&E Guidelines or simply MRP-227, Revision 1-A) provides industry background for the guidelines, lists of RVI components requiring inspection, and the timing for inspections of those components. For each component, the guidelines require a specific type of nondestructive examination (NDE) and give criteria for evaluating inspection results. MRP-227, Revision 1-A provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W, and CE). The document was submitted to the NRC for a formal evaluation and review. The Safety Evaluation Report (SER) was issued on April 25, 2019 [52].

• MRP-228 [53], Inspection Standard for Pressurized Water Reactor Internals - 2018 Update, provides guidance on the qualification/demonstration of the required NDE techniques and other criteria pertaining to the actual performance of the inspections. Reference 53 is the most recent revision of MRP-228 at the time of this report; the revision of MRP-228 in effect at the time of the inspection will be utilized for future inspections.

Additionally, the PWROG has developed WCAP-17096-NP, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements for the MRP-227 inspections. In 2016, the PWROG reissued WCAP-17096-NP as WCAP-17096-NP-A Revision 2 after receiving NRC staff review and approval [11].

2.2 CATAWBA REACTOR VESSEL INTERNALS The RVI for Catawba are integral with the reactor coolant system (RCS) of a Westinghouse four-loop nuclear steam supply system (NSSS). Illustrations of typical RVI are provided in Figure A-1 through Figure A-12. As described in the License Renewal Application (LRA) [12], the RVI consist of the upper core support structure, the lower core support structure, and the in-core instrumentation support structure; each of these major component assemblies has a distinct purpose. The flux thimble tubes, although not part of the RVI, are being addressed because of their inclusion in MRP-227, Revision 1-A. The flux thimble tubes extend from the penetrations in the reactor vessel lower head to the seal table.

The upper core support structure consists of the upper support plate assembly, the upper core plate, upper head injection (UHI) support columns, and the control rod guide tube assemblies. The UHI support columns establish the spacing between the upper support plate assembly and the upper core plate and are fastened at the top and bottom to these plates. The UHI support columns transmit the mechanical loadings between the two plates and serve the supplementary function of supporting thermocouple guide tubes.

They position the upper core plate and upper support plate assembly which act as the boundaries for the flow plenum at the outlet of the core. The control rod guide tube assemblies shield and guide the control rod drive shafts and control rods. The guide tubes are fastened to the upper support plate and are restrained by pins in the upper core plate for proper orientation and support. The upper core support WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 2-3 structure is positioned in its proper orientation with respect to the lower support structure by flat-sided pins pressed into the core barrel which in turn engage in slots in the upper core plate. The upper core support structure is restrained from any axial movements by a large circumferential spring which rests between the upper barrel flange and the upper core support assembly. The entire upper core support structure is removed as a unit during refueling operations to permit access to the fuel assemblies [5].

Vertical loads from dead weight, earthquake acceleration, hydraulic loads, and fuel assembly preload are transmitted through the upper core plate via the support columns to the upper support plate assembly, and then into the reactor vessel head. Transverse loads from coolant crossflow, earthquake acceleration, and possible vibrations are distributed by the support columns to the upper support plate and upper core plate.

The upper support plate is particularly stiff to minimize deflection [5].

The lower core support structure consists of: the core barrel, the core baffle, the lower core plate, support columns, neutron shield pads, and the core support (which is welded to the core barrel). The lower core support structure is supported at its upper flange by a ledge in the reactor vessel head flange and its lower end is restrained from transverse movement by a radial support system attached to the reactor vessel wall.

Within the core barrel are an axial baffle and a lower core plate, both of which are attached to the core barrel wall and form the enclosure periphery of the assembled core. The lower core support structure and, principally, the core barrel serve to provide passageways and control for the coolant flow. The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies. The lower core plate is a member through which the necessary flow distribution holes for each fuel assembly are machined. Fuel assembly locating pins (two (2) for each assembly) are also inserted into this plate. Columns are placed between this plate and the core support of the core barrel to provide stiffness and to transmit the core load to the core support [5].

The purpose of the lower core support structure is to form a peripheral enclosure of the core including core baffles and a bottom flow distribution plate for efficient flow distribution, provide neutron shielding by means of the neutron shield pads, and provide structural support to withstand transverse loadings from coolant crossflow and other design conditions. The lower core support structure also provides structural support for vertical loads from the fuel, hydraulic forces, control rod dynamics, and other design loadings

[5]. The lower core support structure remains in place in the reactor vessel during most refueling operations. Typically, it is only removed to perform scheduled reactor vessel inspections or inspection of the lower core support structure itself.

Vertically downward loads from weight, fuel assembly preload, control rod dynamic loading, hydraulic loads, and earthquake acceleration are carried by the lower core plate partially into the lower core plate support flange on the core barrel shell and partially through the lower support columns to the core support, and then through the core barrel shell to the core barrel flange supported by the vessel head flange. Transverse loads from earthquake acceleration, coolant crossflow, and vibration are carried by the core barrel shell and distributed between the lower radial support and the vessel flange. Transverse loads of the fuel assemblies are transmitted to the core barrel shell by direct connection of the lower core plate to the barrel wall and by upper core plate alignment pins, which are welded into the core barrel [5].

The purpose of the in-core instrumentation support structure is to provide structural support for the bottom-mounted in-core instrumentation (flux thimbles and thermocouples).

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Westinghouse Non-Proprietary Class 3 2-4 2.3 CATAWBA LICENSE RENEWAL Duke Energy submitted an LRA for a renewed operating license for Catawba [12]. In the SER (NUREG-1772) [1] for the LRA, the NRC concluded that the applicant has demonstrated that the aging effects associated with the RVI will be adequately managed, so there is reasonable assurance that these components will perform their intended function(s) consistent with the CLB throughout the period of extended operation, as required by 10 CFR 54.21(a)(3).

A listing of the Catawba RVI components and subcomponents that are subject to AMP requirements is in Table 3.1-1 of the LRA. In the SER, the NRC concluded that the Duke LRA adequately identified the RVI system structures and components that are subject to an aging management review (AMR), as required by 10 CFR 54.21(a)(1). A listing of the Catawba Unit 1 and Unit 2 reactor vessel internals components and subcomponents already reviewed by the NRC is included in Table B-1. Included in the SER, Appendix D is Commitment 14 associated with the Reactor Vessel Internals Inspection.

The existing inspection commitments are contained in Section 18.2.22 of the Catawba UFSAR. The UFSAR section contains an allowance that permits Duke Energy to modify or eliminate these inspections if plant-specific justification is provided to demonstrate the basis for the modification or elimination.

Additionally, as part of its license renewal, Duke Energy stated they would participate in industry activities associated with RVI-related issues and that the Catawba RVI Program is subject to future enhancements as the industrys understanding of degradation continues to improve. The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI. The results of the industry recommended inspections serve as the basis for identifying any augmented inspections that are required to complete the Catawba RVI Program. In a letter dated June 16, 2010 [32], Duke Energy notified the NRC of its intent to revise its commitments for RVI inspections from those that currently exist in Catawbas UFSAR to the inspection guidelines provided by MRP-227 as approved by the NRC [32]. Since the June 16, 2010 letter [32], MRP-227 has been revised to Revision 1 and has received NRC approval through the MRP-227 Revision 1 SER issued on April 25, 2019 [52]. Therefore, this CNS RVI AMP and Inspection Plan, as documented herein, is based on MRP-227, Revision 1-A. Once the Catawba RVI AMP and Inspection Plan is approved by the NRC, the Catawba UFSAR will be updated as required.

The original license renewal commitment did not include the submittal of an inspection plan. The 2010 letter [32] did not provide a schedule date for implementation of the MRP-227 guidelines. However, in letter dated March 19, 2014 [28], Duke Energy notified the NRC of its intent to submit a plant-specific RVI inspection plan for Catawba to implement MRP-227-A (or latest NRC approved revision, currently MRP-227, Revision 1-A) no later than 2 years before the initial inspection. Within [28], Duke Energy also notified the NRC that the expected inspection plan submittal date for Catawba is Fall 2022. The expected initial inspection dates are Fall 2024 and Fall 2025 for Catawba Units 1 and 2, respectively, but may be subject to change.

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Westinghouse Non-Proprietary Class 3 3-1 3 PROGRAM OWNER The successful implementation and comprehensive long-term management of the Catawba RVI program will require the integration of Duke Energy organizations, corporately and at Catawba, and interaction with multiple industry organizations including, but not limited to, the ASME, MRP, NRC, and PWROG.

Duke Energy will maintain cognizance of industry activities related to PWR internals inspection and aging management and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

The overall responsibility for administration of the RVI program is Catawba senior management.

Roles and responsibilities for establishing, maintaining, and implementing the Catawba RVI Program are established in the applicable Duke Energy administrative and program procedures [31].

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Westinghouse Non-Proprietary Class 3 4-1 4 DESCRIPTION OF CATAWBA REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS The U.S. nuclear industry, through the combined efforts of utilities, vendors, and independent consultants, has defined a generic guideline to assist utilities in developing RVI plant-specific programs based on inspection and evaluation. The intent of the Catawba RVI program is to ensure the long-term integrity and safe operation of the RVI components. Catawba developed this AMP and Inspection Plan in conformance with NUREG-1801, Revision 2 as updated by LR-ISG-2011-04 [6 and 38] and MRP-227, Revision 1-A

[51].

This RVI program utilizes a combination of prevention, mitigation, and condition monitoring. Where applicable, credit is taken for existing programs, such as water chemistry [14] and inspections prescribed by the ASME Section XI ISI Program [39, 40, 41, and 5] and past and future mitigation projects, such as control rod guide tube support pin replacement and the bottom-mounted instrumentation thimble tube program. These existing programs are augmented with the inspections and evaluations recommended by MRP-227, Revision 1-A.

Aging degradation mechanisms that impact internals have been identified and documented in the LRA submitted by Duke Energy [12]. The overall outcome of the reviews and the additional work performed by the industry, as summarized in MRP-227, Revision 1-A, is to provide appropriate augmented inspections for RVI components to provide early detection of the degradation mechanisms of concern.

Therefore, this AMP and Inspection Plan is consistent with the existing Catawba RVI AMR methodology and the additional industry work summarized in MRP-227, Revision 1-A. All sources are consistent and address concerns about component degradation resulting from the following eight (8) material aging degradation mechanisms identified as affecting RVI:

• Stress Corrosion Cracking Stress corrosion cracking (SCC) refers to local, nonductile cracking of a material due to a combination of tensile stress, environment, and metallurgical properties. The actual mechanism that causes SCC involves a complex interaction of environmental and metallurgical factors. The aging effect is cracking.

• Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only in highly irradiated components. The aging effect is cracking.

• Wear Wear is caused by the relative motion between adjacent surfaces, with the extent determined by the relative properties of the adjacent materials and their surface condition. The aging effect is loss of material.

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• Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.

Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates. When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue. The aging effects of low-cycle fatigue and high-cycle fatigue are additive.

Fatigue crack initiation and growth resistance are governed by a number of material, structural, and environmental factors such as stress range, loading frequency, surface condition, and presence of deleterious chemical species. Cracks typically initiate at local geometric stress concentrations such as notches, surface defects, and structural discontinuities. The aging effect is cracking.

• Thermal Aging Embrittlement Thermal aging embrittlement (TE) is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardening (PH) stainless steel to high in-service temperatures, which can result in an increase in tensile strength, a decrease in ductility, and a loss of fracture toughness. Some degree of thermal aging embrittlement can also occur at normal operating temperatures for CASS and PH stainless steel internals. CASS components have a duplex microstructure and are particularly susceptible to this mechanism. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

• Irradiation Embrittlement Irradiation embrittlement (IE) is also referred to as neutron embrittlement. When exposed to high-energy neutrons, the mechanical properties of stainless steel and nickel-based alloys can be changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and a loss of fracture toughness. The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

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• Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters of irradiation-produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling

(> 5% by volume) has been correlated with extremely low fracture toughness values. Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes within in-core instrumentation tubes that are fabricated from zirconium alloys. While the initial aging effect is dimensional change and distortion, severe void swelling may result in cracking under stress.

• Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation-Enhanced Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, as seen in PWR internals.

Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (< 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />) at PWR internals temperatures.

Creep (or more precisely, secondary creep) is a slow, time- and temperature-dependent, plastic deformation of materials that can occur when the material is subjected to stress levels below the yield strength (elastic limit). Creep occurs at elevated temperatures where continuous deformation takes place under constant stress. Secondary creep in austenitic stainless steel is associated with temperatures higher than those relevant to PWR internals even after taking into account gamma heating. However, irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress relaxation (ISR) is an athermal process that depends on the neutron fluence and stress, and it can also be affected by void swelling should it occur. The aging effect is a loss of mechanical closure integrity (or preload) that can lead to unanticipated loading that, in turn, may eventually cause subsequent degradation by fatigue or wear and result in cracking.

The program to manage the aging of the Catawba RVI, which will include this AMP and Inspection Plan once it is approved by the NRC, incorporates programs and activities that are credited for managing the aging effects produced by the aging degradation mechanisms listed above.

4.1 SUPPORTING CATAWBA PROGRAMS AND ACTIVITIES Catawba has a number of programs and activities related to the aging management of the RVI; these include:

• Chemistry Control Program

• ASME Section XI ISI Program

• Bottom-Mounted Instrumentation Thimble Tube Inspection Program WCAP-18496-NP September 2020 Revision 0

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• Control Rod Guide Tube Support Pin Replacement Project

• Power Uprating Project Brief descriptions of the programs and activities are included in the following subsections.

4.1.1 Chemistry Control Program The Catawba Chemistry Control Program is an existing program that provides activities for monitoring and controlling the chemical environments of the Catawba primary cycle systems such that aging effects of system components are minimized. This program manages the aging effects of cracking and loss of material. The program mitigates damage caused by corrosion and SCC and other aging mechanisms. This program includes provisions specified for the verification of proper chemistry control and aging management, such that the intended functions of plant components will be maintained during the period of extended operation for Catawba.

The Catawba Chemistry Control Program includes periodic sampling of primary water for the known detrimental contaminants specified in the EPRI PWR water chemistry guidelines [15] to maintain their concentrations below levels known to result in loss of material or cracking. Sampling frequencies and action limits for each control parameter are defined in Catawba-specific procedures [14].

Catawba follows the guidance set forth in the EPRI PWR Primary Water Chemistry Guidelines [15]. The limits imposed by the Catawba Program meet the intent of the industry standard for addressing primary water chemistry [14].

4.1.2 ASME Section XI Program The Catawba ASME Section XI Program [39, 40, and 41] is an existing program that includes examinations of the Reactor Vessel core support structure components in accordance with ASME Section XI, Subsection IWB-2500. Core support structures are examined using visual VT-3 examination methods each interval (Examination Category B-N-3). Table 4-9 in MRP-227, Revision 1-A lists the existing programs that are credited for aging management in the Westinghouse-design plants. Many of these component items are considered core support structures that are typically examined during the 10-year in-service inspection [41]. Table 4-9 considers the Code requirements for visual VT-3 sufficient to monitor the applicable aging effects for the Existing Programs components, with the exception of the clevis bearing stellite wear surface, clevis insert bolts, lower core plate, XL lower core plate (not applicable to Catawba), and upper core plate alignment pins. The Code required exam for the clevis bearing stellite wear surface and clevis insert bolts is supplemented by TB-14-5 [33], and the Code required exam for the lower core plate and upper core plate alignment pins is supplemented by TB-16-4 [86]. Catawba will follow the Code requirements for visual VT-3 examination and will follow TB-14-5 for the clevis bearing stellite wear surface and clevis insert bolts. TB-16-4 is not applicable to Catawba.

4.1.3 Bottom-Mounted Instrumentation Thimble Tube Inspection Flux thimble tubes are long, slender, stainless steel tubes that are seal welded at one end with flux thimble tube plugs, which pass through the vessel penetration, through the lower internals assembly, and finally extend to the top of the fuel assembly. The bottom-mounted instrumentation (BMI) column assemblies WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 4-5 provide a path for the flux thimbles into the core from the bottom of the vessel and protect the flux thimbles during operation of the reactor. The flux thimble provides a path for the neutron flux detector into the core and is subject to reactor coolant pressure on the outside and containment pressure on the inside.

The Catawba Bottom-Mounted Instrumentation Thimble Tube Inspection Program is a program that manages loss of material due to wear of the flux thimble tube materials [26] and [27]. It implements the recommendations of NRC Bulletin 88-09 [16] that a thimble tube wear inspection procedure be established and maintained for Westinghouse-supplied reactors that use bottom-mounted flux thimble tube instrumentation. The program utilizes an inspection methodology, such as eddy current testing (ET),

to inspect the flux thimble tubes on a periodic frequency and to monitor wall thinning and predict when tubes would require repair or replacement. The program implements a wall thickness trending report.

The BMI Thimble Tube Inspection Program establishes appropriate acceptance criteria (percent through-wall wear), based on industry guidance and includes margin to allow for factors such as instrument uncertainty, uncertainties in wear scar geometry, and other potential inaccuracies, as applicable, to the inspection methodology [54]. Table 4-9 in MRP-227, Revision 1-A lists the Existing examinations that are credited for aging management in the Westinghouse-design plants. Included in Table 4-9 are the Bottom Mounted Instrumentation System Flux Thimble Tubes. For this item, the Eddy current examination as defined in the plants response to NRC Bulletin 88-09 [16] is considered sufficient to monitor for the applicable aging effect.

4.1.4 Control Rod Guide Tube Support Pin Replacement Project The control rod guide tube support pins are used to align the bottom of the control rod guide tube assembly into the top of the upper core plate.

Because of support pin cracking experienced in the industry, the original Catawba support pins, at both Unit 1 and Unit 2, were changed to a replacement material prior to startup. The new support pins were fabricated from INCONEL1 Alloy X-750, but with a modified heat treatment [45]. Support pins made of this material with the associated heat treatments continued to be susceptible to primary water SCC (PWSCC) and proved likely to fail during the lifetime of a nuclear power plant. Westinghouse developed an improved support pin design made of Type 316 stainless steel (SS) material with a fabrication technique that significantly reduced the susceptibility to PWSCC while maintaining the fatigue and wear requirements necessary to support continued uninterrupted service [17 and 23]. In response to industry concern, the Alloy X-750 support pins were replaced with Type 316 SS support pins at Catawba Unit 1 in Fall 2000 [87] and at Catawba Unit 2 in Fall 2001 [88]; the replacement support pins utilized improved materials that support the proactive management of aging in RVI components.

Support pins fabricated from Type 316 SS are not susceptible to PWSCC which was the primary failure mechanism for Alloy X-750 support pins. MRP-191, Revision 1 categorized the guide tube support pins as Category A, which is assigned to components for which the aging effects are below the screening criteria or for which again degradation significance is minimal [36]. Section 4.5 of MRP-227, Revision 1-1 INCONEL is a trademark or registered trademark of Special Metals; a Precision Castparts Corp. company. Other names may be trademarks of their respective owners.

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Westinghouse Non-Proprietary Class 3 4-6 A describes this categorization of the 316 SS support pins and defines the Type 316 SS support pins as a no additional measures component [51]. Thus, the Type 316 SS support pins are not included in Table 4-9 of MRP-227, Revision 1-A and do not require a plant-specific aging management program.

Therefore, no additional inspections are required by the supplier or per MRP-227, Revision 1-A, and the support pins should remain functional for the period of extended operation.

4.1.5 Power Uprating Project for Catawba Unit 1 Catawba Unit 1 was originally licensed to 3411 megawatts thermal (MWt) and sought to increase core power to 3469 MWt, through the use of more accurate feedwater flow measurement instrumentation [29].

During the review of [29], the NRC provided Requests for Additional Information (RAIs), which Duke Energy responded to within [56], [57], [58], and [59]. Performance of this power uprate was approved by the NRC per Catawba Nuclear Station Units 1 and 2 License Amendments 281 and 277 and resulted in greater power generation of electricity for Catawba Unit 1 [30]. The power uprate design evaluation [29]

ensured operation of Catawba Unit 1 remained consistent with safety-related analyses and remained below design basis limits. A power uprate was not performed for Catawba Unit 2 because the uprate would require the replacement of its steam generators [29]. The submittals referenced are docketed for both Catawba Unit 1 and Unit 2 since the technical specifications are common to both units.

The evaluation of the power uprate performed at Catawba Unit 1 and detailed description of the changes are available in plant records [29]. The NRC staff reviewed the licensee's evaluation of the impact of the license amendment request (LAR) on the structural integrity assessments for the RVI. The NRC staff determined within [30] that the licensee's RVI evaluation considering the effect of the LAR was acceptable because: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

4.2 INDUSTRY PROGRAMS 4.2.1 WCAP-14577, Aging Management for Reactor Internals The WOG topical report WCAP-14577 [8] contains a technical evaluation of aging degradation mechanisms and aging effects for Westinghouse RVI components. The WOG sent the report to the NRC staff to demonstrate that WOG member plant owners that subscribed to the WCAP could adequately manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies of the WCAP to develop plant-specific AMPs.

Duke Energy did not use WCAP-14577 as a reference to complete the AMR for the Catawba internals.

However, the NRC referred to WCAP-14577 in their review of the Catawba LRA and requested additional information where the Catawba LRA was inconsistent with WCAP-14577 [1].

4.2.2 MRP-227, Revision 1-A, Reactor Internals Inspection and Evaluation Guidelines MRP-227, Revision 0 [3], as discussed in Section 2, was developed by a team of industry experts including: utility representatives, NSSS vendors, independent consultants, and international committee WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 4-7 representatives who reviewed available data and industry experience on materials aging. The objective of the group was to develop a consistent, systematic approach for identifying and prioritizing inspection and evaluation requirements for RVI. MRP-227 has since been updated to MRP-227, Revision 1-A [51],

which has been accepted by the NRC [89], see Section 6.1.3.

MRP-227, Revision 1-A Reactor Vessel Internals Component Categorizations MRP-227, Revision 1-A used a screening and ranking process to aid in the identification of required inspections for specific RVI components. MRP-227, Revision 1-A credits existing component inspections, when they were deemed adequate, as a result of detailed expert panel assessments conducted in conjunction with the development of the industry document. Through the elements of the process, the reactor internals for all currently licensed and operating PWR designs in the U.S. were evaluated in the MRP Program and appropriate inspection, evaluation, and implementation requirements for reactor internals were defined.

Based on the completed evaluations, the RVI components are categorized within MRP-227, Revision 1-A as Primary components, Expansion components, Existing Programs components, or No Additional Measures components, described as follows:

• Primary Those PWR internals that are highly susceptible to the effects of at least one of the eight (8) aging mechanisms were placed in the Primary group. The aging management requirements are intended to provide reasonable assurance of the continued functionality of Primary components and to predict future behavior of Expansion components as described in these I&E guidelines.

Where little to no service degradation has been experienced to date and/or service degradation is not expected solely based on the aging mechanism, a sampling strategy for primary components is specified. The Primary group also includes components which have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.

• Expansion Those PWR internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which engineering evaluations and safety assessments have shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components will depend on the findings from the examinations of the Primary components at individual plants. In this regard, the Expansion group also consists of those components for which an increased scope of the Primary component sample is specified based on degradation detected in the Primary sample (i.e.,

increased sampling of a Primary component).

• Existing Programs Those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which generic and plant-specific existing AMP elements (e.g., ASME B&PV WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 4-8 Code Section XI [2]) are capable of managing those effects, were placed in the Existing Programs group.

• No Additional Measures Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of the failure modes, effects, and criticality analysis (FMECA) and the engineering evaluations and safety assessments. No further action is required by these guidelines for managing the aging of the No Additional Measures components.

The categorization and analysis used in the development of MRP-227, Revision 1-A are not intended to supersede any ASME B&PV Code Section XI requirements. Any components that are classified as core support structures, as defined in ASME B&PV Code Section XI IWA-9000, and covered by Table IWB-2500-1 Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a or plant-specific licensing documentation.

The applicability of the MRP-227, Revision 1-A guidelines to Catawba Unit 1 and Unit 2 is described in detail within Section 6.1.

4.2.3 WCAP-17451-P, Reactor Internals Guide Tube Wear The PWROG developed a tool to facilitate prediction of upper internals control rod guide tube assembly guide card and lower guide tube continuous guidance wear. An initial inspection schedule based on the various guide tube designs for the utilities participating in this program and acceptance criteria was then established. WCAP-17451-P, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections [60], documents the guide plates (cards) initial inspection schedule and acceptance criteria for Westinghouse NSSS designed plants. The examination method/frequency for the guide plates (cards) is provided in WCAP-17451-P, Revision 1 [25] with the modified requirements of interim guidance provided in EPRI letter MRP 2018-007 [61] and PWROG letter OG-18-46 [62], as seen in Table 4-3 of MRP-227, Revision 1-A. The latest revision of WCAP-17451-P, Revision 2 [60], includes the interim guidance from MRP 2018-007 [61] and OG-18-46 [62]. Therefore, Duke Energy has chosen to follow Revision 2 for guide plate (card) inspections. WCAP-17451-P, Revision 2 was transmitted to the NRC in PWROG letter OG-19-197 [64].

Catawba Unit 1 and Unit 2 are 4-loop plants with a 17x17A guide tube design that has used ion nitride rod cluster control assemblies (RCCAs). Recent OE at U.S. Westinghouse NSSS plants that have 17x17 A or 17x17 AS style guide tubes and have switched to ion nitride RCCAs indicates that the rate of guide card wear has outpaced the predications in WCAP-17451-P, Revision 1, as stated in NSAL-17-1 [49].

This issue was determined to have a potential nuclear safety consequence and therefore was reported to the NRC as a defect, pursuant to 10 CFR Part 21 [65]. WCAP-17451-P, Revision 2 includes updated Needed guidance for addressing accelerated guide card wear and PWROG letter OG-18-276 [63]

transmitted this Needed guidance to the PWROG members. Catawba Unit 1 and Unit 2 will therefore follow the inspection and evaluation guidance within WCAP-17451-P, Revision 2.

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Westinghouse Non-Proprietary Class 3 4-9 All 53 guide tubes at rodded locations at Catawba Unit 1 were inspected during the Fall 2018 outage [34].

This inspection was performed per WCAP-17451-P, Revision 2 [60]. Two of the worst worn guide tubes were relocated to unrodded locations in the upper internals and were replaced with unworn spare guide tubes previously in the unrodded positions. All other relevant indications identified were dispositioned with no operability concerns. Additionally, Catawba Unit 1 replaced the ion nitride RCCAs with chrome-plated RCCAs in Spring 2020. Catawba is following the guidance in WCAP-17451-P, Revision 2 for future management of guide card and lower guide tube continuous guidance wear at Catawba Unit 1.

All 53 guide tubes at rodded locations at Catawba Unit 2 were inspected during the Fall 2016 outage [35].

The inspections were originally performed using WCAP-17451-P, Revision 1 criteria. A total of eight (8) of the worst worn guide tubes were relocated to unrodded locations in the upper internals and were replaced with unworn spare guide tubes previously in the unrodded locations, during the Fall 2016 and Spring 2018 outages. The guide card wear measurements from the Fall 2016 outage were reanalyzed with guidance from WCAP-14751-P, Revision 2 within Revision 2 of WCAP-18198-P. Additionally, Catawba replaced the ion nitride RCCAs with chrome-plated RCCAs in Fall 2019. Catawba is following the guidance in WCAP-17451-P, Revision 2 for future management of guide card and lower guide tube continuous guidance wear at Catawba Unit 2.

4.2.4 Baffle-Former Bolt Degradation Recently, a larger-than-expected number of degraded baffle-former bolts were discovered in 4-loop downflow plants through MRP-227 inspection requirements and reactionary inspections at like plants.

The pattern of these degraded bolts was also concentrated, or clustered, more than anticipated based on OE gained from previous analyses and inspections. As a result, Westinghouse performed a 10 CFR 21 evaluation, the results of which are discussed in NSAL-16-1 [46]. NSAL-16-1 also provided guidance for affected utilities by grouping plants based on their susceptibility to baffle former bolt degradation and offered recommendations on re-inspection techniques and intervals. Interim guidance has been developed and published by the MRP for the Tier 1 plants in MRP 2016-021 [48] and for all of the NSAL-16-1 plants in MRP 2017-009 [47], respectively. The interim guidance within MRP 2016-021 and MRP 2017-009 has been incorporated into MRP-227, Revision 1-A, Table 4-3.

Both Catawba Unit 1 and Unit 2 are upflow configuration plants and therefore fall into the Tier 4 plant category described in NSAL-16-1. An upflow configuration has been shown to reduce the incidence of baffle jetting damage to fuel and to reduce the bolt loads due to pressure differentials across the baffle under both normal operating and expected faulted conditions. Based on MRP-227, Revision 1-A and interim guidance MRP 2017-009 [47], Catawba will perform a baseline volumetric (UT) examination no later than 35 effective full-power years (EFPY). This corresponds to the Fall 2024 outage for Catawba Unit 1 and the Fall 2025 outage for Catawba Unit 2.

In accordance with MRP 2018-002 [76], if Catawba discovers significant baffle-former bolt clustering as defined in MRP 2018-002, Catawba will perform a one-time visual (VT-3) examination of barrel-former bolts within three (3) fuel cycles. The VT-3 examination coverage will include the barrel-former bolts adjacent to the large clusters of baffle-former bolts with unacceptable indications, as defined in MRP 2018-002.

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Westinghouse Non-Proprietary Class 3 4-10 4.2.5 WCAP-17096, Reactor Internals Acceptance Criteria Methodology and Data Requirements The industry, through various cooperative efforts, is working to construct a consensus set of tools in line with accepted and proven methodologies to support Generic Aging Lessons Learned (GALL) Revision 2, Element 6, related to acceptance criteria. One of these tools is the PWROG document WCAP-17096-NP-A, Reactor Internals Acceptance Criteria Methodology and Data Requirements [11], which details acceptance criteria methodology for the MRP-227-A Primary and Expansion components. Revision 2 of WCAP-17096-NP was transmitted to the NRC on May 19, 2010 for review. The NRC issued a draft SE on November 12, 2015 and received comments from the industry on January 28, 2016. The NRC accepted WCAP-17096-NP, Revision 2 on May 3, 2016 allowing the issuance of WCAP-17096-NP-A

[11].

WCAP-17096-NP-A and the guidance within PWROG-17071-NP [79], provide acceptable methodology to evaluate inspection findings during MRP-227 examinations. The interim guidance within PWROG-17071-NP focuses on areas where the currently approved acceptance criteria guidance within WCAP-17096-NP-A is inconsistent with other industry guidance that has been issued since the NRC approval of WCAP-17096-NP-A. PWROG-17071-NP was issued to the Pressurized Water Reactor Owners Group members within OG-18-61 [80].

The status of WCAP-17096-NP-A is monitored through direct Duke Energy cognizance of industry (including PWROG) activities related to PWR internals inspection and aging management. Revision 3 of WCAP-17096-NP [81] has been sent to the NRC for review and acceptance within OG-19-164 [82].

WCAP-17096-NP, Revision 3 includes the interim guidance from PWROG-17071-NP and updates to the acceptance criteria guidance for MRP-227, Revision 1 as well as applicability to initial license renewal (60-year operating license) and subsequent license renewal (80-year operating license) based on current industry knowledge.

4.2.6 Core Barrel Operating Experience During Spring 2018 inspections, one (1) CE-designed plant identified crack-like surface indications at the core support barrel (CSB) assembly welds, specifically one (1) vertically oriented indication at the middle girth weld and 45 indications adjacent to the middle axial weld. The majority of the middle axial weld indications were oriented perpendicular to the weld, circumferential to the barrel. A supplemental volumetric examination to characterize the indications was subsequently performed. The examination confirmed that the visually identified indications did not extend through the CSB thickness. These flaws were located in material with neutron dose levels high enough for potential IASCC, but the degradation mechanism has not been confirmed. The potential for cracking of CE-designed CSBs and Westinghouse-designed core barrels was acknowledged in the previous I&E guidelines; however, the observed condition was inconsistent with expectations for number, location, and orientation of the indications. This OE was communicated to the PWR fleet in MRP 2018-028 [83].

NEI 03-08 Good Practice interim guidance for MRP-227-A was developed within MRP 2019-009 [84].

Based on the evaluations conducted in Attachment 1 and Attachment 2 of MRP 2019-009, it is concluded that age-related cracking at axial welds is not likely to be any more prevalent or safety significant than cracking at circumferential welds. Although the MRP-227, Revision 1-A reactor internals inspection and WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 4-11 evaluation guidance as written may allow for an inspection to overlook circumferential cracking in an axial weld, the monitoring of girth welds remains a reasonable surrogate for identification of age-related cracking of the core barrel prior to any significant degradation which could impact plant safety from the standpoint of shutdown capability and core damage. The EVT-1 inspection continues to be an acceptable visual examination method for detecting cracking in the barrel welds. The guidance within MRP 2019-009 is effective at all Westinghouse-designed and CE-designed US PWR units as of February 1, 2020.

Both Catawba Unit 1 and Unit 2 fall into Group 2 of MRP 2019-009 and the "Good Practice" guidance is being evaluated.

4.2.7 Clevis Bearing Stellite Wear Surface and Clevis Insert Bolts - Technical Bulletin TB-14-5 Technical Bulletin TB-14-5 [33] provides a summary of the OE for the clevis bearing stellite wear surface and clevis insert bolts, as well as the root cause findings and the applicability of these findings on Westinghouse and CE pressurized water reactor designs. Technical Bulletin TB-14-5 also reviews the safety implications of the OE and root cause analysis results and provides inspection recommendations for licensees to consider for inclusion as part of their aging management program to address this OE.

Catawba committed to perform a VT-1 of the clevis insert fasteners during the review of the License Renewal Application. Catawba, as stated previously, will follow the recommendations of MRP-227, Revision 1-A, as supplemented by TB-14-5 for the aging management of this component.

4.2.8 Thermal Sleeve Flange Wear Reactor vessel closure head thermal sleeves have a history of wear by several different mechanisms at various locations, as described within TB-07-2 [66]. During the Spring 2014 outage season, an additional thermal sleeve wear mechanism was identified at one plant when a thermal sleeve at a partial-length control rod drive mechanism (CRDM) location fell from the RV closure head during an ISI [66].

Examination of the fallen sleeve showed that the upper flange, which rests inside the CRDM head adapter tube, had worn through. Further inspection of the damaged sleeve and adapter concluded that a mutual wear mechanism existed between the two components. The wear was attributed to flow- and pump-induced vibration.

An Owners Group report, PWROG-16003-P [68], was created to provide the technical basis and acceptance criteria for evaluating thermal sleeve flange wear in response to the OE within TB-07-2 [66].

The evaluation of separated sleeves and flanges within PWROG-16003-P, Revision 0 and Revision 1, concluded that control rod interference with a degraded thermal sleeve was unlikely. However, further operating experience, as described within NSAL-18-1 [67], has shown that Westinghouse NSSS plants that have thermal sleeves in CRDM penetration tubes have the potential for wear of the thermal sleeve flange against the head adapter tube. This wear can lead to consequences that were not previously considered in [66]. In accordance with 10 CFR Part 21, Westinghouse reported the thermal sleeve flange wear issue within NSAL-18-1 as a potential defect in May 2018 [73]. While there have been no reported events of control rods failing to insert into the core when required, Westinghouse reported this issue to the NRC under 10 CFR Part 21 because it had the potential to create a substantial safety hazard. PWROG-16003-P was updated to Revision 2 [68] to account for the OE reported within NSAL-18-1 [67] and was transmitted to the PWROG members within OG-19-101 [69].

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Westinghouse Non-Proprietary Class 3 4-12 NEI 03-08 Needed guidance was developed and approved within MRP 2018-027 [72] by the joint MRP/PWROG utility members and PMMP-EC/PWROG-EC members to provide immediate guidance for the Table 1 and Table 2 plants as defined in the NSAL 18-1 [67]. Table 1 of NSAL-18-1 includes plants with higher susceptibility of thermal sleeve flange wear and includes both Catawba Unit 1 and Unit 2.

The latest OE as described within NSAL-20-1 [91] has shown that Westinghouse NSSS plants that operate in a T-cold configuration, and that have CRDM thermal sleeves with a collar below the flange, are potentially susceptible to cracking and separation of the flange from the sleeve. This separated condition, when combined with the type of flange wear discussed in NSAL-18-1 [67], can potentially impede control rod movement and RCCA insertability. In December 2019, in accordance with 10 CFR Part 21, Westinghouse reported this issue as a potential defect due to the possibility for this type of degradation to impact control rod insertion and ultimately plant shutdown [92]. Although this OE involved a new thermal sleeve failure mechanism, the safety significance of this issue is associated with the conditions discussed in NSAL-18-1 [67]. As a result, the recommendations captured in NSAL-20-1 [91] should be considered in addition to those provided in NSAL-18-1 [67]. The PWROG transmitted NEI 03-08 Needed and Good Practice guidance regarding NSAL-20-1 [91] to the industry within letter OG 113 [97]. Catawba Unit 1 and Unit 2 both fall into the Table 1 design category listed in NSAL-20-1 [91].

Catawba is evaluating the OE and intends to comply with the recommendations in NSAL-20-1 and the Needed guidance within OG-20-113 [97] for the aging management of the thermal sleeves.

Duke Energy performed a thermal sleeve flange wear inspection at Catawba Unit 2 during the Spring 2018 refueling outage [70] in accordance with TB-07-2 [66] using the evaluation methodology and acceptance criteria within PWROG-16003-P, Revision 1 [85]. NSAL-18-1 [67] was not published at this time. This inspection indicated that no rodded locations exceeded the generic separation acceptance criteria established in PWROG-16003-P. Catawba is using plant-specific wear rates and acceptance criteria based on the methodology in PWROG-16003-P for future management of thermal sleeve flange wear at Catawba Unit 2.

Duke Energy performed a thermal sleeve flange wear inspection at Catawba Unit 1 during the Fall 2018 refueling outage [71] in accordance with MRP 2018-027 [72] and NSAL-18-1 [67] using the methodology and acceptance criteria within PWROG-16003-P, Revision 1 [85]. This inspection indicated that no rodded locations exceeded the generic separation acceptance criteria established in PWROG-16003-P. Catawba is using plant-specific wear rates and acceptance criteria based on the methodology in PWROG-16003-P for future management of thermal sleeve flange wear at Catawba Unit 1.

4.2.9 On-Going Industry Programs and NEI 03-08 Guidelines As part of its license renewal, Duke Energy stated they would participate in industry activities associated with RVI-related issues and that the Catawba RVI Program is subject to future enhancements as the industrys understanding of degradation continues to improve. The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI. The results of the industry-recommended inspections serve as the basis for identifying any augmented inspections that are required.

Several of the industry work products developed in support of the aging management of PWR internals are issued by Issue Programs (including MRP and PWROG) under the implementation protocol of WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 4-13 NEI 03-08 [13]. Included work products are MRP-227, MRP-228, WCAP-17451-P, and any interim guidance associated with these documents. Appendix B to NEI 03-08, Implementation Protocol, defines the processes and expectations for implementing industry guidance issued under the Materials Initiative, and requires that Issue Programs identify the specific implementation category for requirements identified by guideline-type work products. While Catawba is basing the AMP and Inspection Plan for RVI on the scope defined in MRP-227, Revision 1-A, Catawba will also implement work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [42], including later revisions and interim guidance to the work products listed above. A failure to meet a Needed or a Mandatory requirement is a deviation from the guidelines and a written justification for the deviation must be prepared and approved as described in Appendix B to NEI 03-08 and Duke Energy administrative procedures, including notification to the NRC for information only.

The U.S. industry, through both the EPRI/MRP and the PWROG, continues to sponsor activities related to RVI aging management. Duke Energy will maintain cognizance of industry activities related to PWR internals inspection and aging management and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices [42].

4.3

SUMMARY

This section contains pertinent Catawba and industry programs and activities used for the development and implementation of MRP-227, Revision 1-A, the Catawba RVI AMP and Inspection Plan, and the Catawba RVI Program.

The augmented inspections described in this document, as summarized in Appendix C, combined with the ASME Code Section XI ISI Program inspections, existing Catawba Programs, and use of OE, provide reasonable assurance that the reactor internals at Catawba will continue to perform their intended functions throughout the period of extended operation.

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Westinghouse Non-Proprietary Class 3 5-1 5 CATAWBA NUCLEAR STATION UNIT 1 AND UNIT 2 REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES Based on Duke Energys revised commitment, the Catawba RVI AMP and Inspection Plan is credited for aging management of RVI components for the following eight (8) aging degradation mechanisms and their associated effects:

• Stress corrosion cracking

• Irradiation-assisted stress corrosion cracking

• Wear (loss of material)

• Fatigue (cracking)

• Thermal aging embrittlement (reduction in fracture toughness)

• Irradiation embrittlement (reduction in fracture toughness)

• Void swelling and irradiation growth (distortion)

• Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep (loss of preload or loss of mechanical closure integrity)

The attributes of the Catawba RVI Program and compliance with NUREG-1801 (GALL Report),Section XI.M16A, "PWR Vessel Internals" [6], as updated via LR-ISG-2011-04 [38] are described in this section.

The Catawba RVI Program is aligned to meet the requirements of MRP-227, Revision 1-A in addition to complying with the GALL and LR-ISG-2011-04 [38]. The GALL identifies ten (10) attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the ten (10) elements of the GALL. The GALL Revision 2 [6] and LR-ISG-2011-04 [38] reference MRP-227-A [24]. MRP-227, Revision 1-A does not alter the 10 elements of the GALL, and as the latest NRC-approved version of the RVI I&E guidelines, MRP-227, Revision 1-A can be used in place of MRP-227-A to satisfy these elements.

This AMP and Inspection Plan is consistent with that process for meeting the ten GALL attributes, considers the augmented inspections identified in MRP-227, Revision 1-A, and fully meets the requirements of the current commitment. Specific details of the Catawba RVI Program are summarized in the following subsections.

5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM GALL Report AMP Element Description The scope of the program includes all RVI components based on the plants applicable nuclear steam supply system design. The scope of the program applies the methodology and guidance in MRP-227-A, which provides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by Babcock & Wilcox (B&W), Combustion Engineering (CE), and Westinghouse. The scope of components considered for inspection in MRP-227-A includes core support structures, those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code,Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 5-2 1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227-A.

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are managed in accordance with an applicants AMP that corresponds to GALL AMP XI.M1, ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD. [38]

Catawba Nuclear Station Unit 1 and Unit 2 Program Scope The Catawba RVI components consist of the upper core support structure, the lower core support structure, and the in-core instrumentation support structure, where each of these major components has a distinct purpose. The flux thimble tubes, although not part of the RVI, are being addressed because of their inclusion in MRP-227-A and MRP-227, Revision 1-A. The flux thimble tubes extend from the penetrations on the reactor vessel lower head up to the seal table. Additional RVI details are provided in the Catawba LRA [12]and UFSAR [5].

The Catawba RVI subcomponents that are subject to AMP requirements were provided in Table 3.1-1 in the Catawba LRA [12]. The table listed each subcomponents intended function(s) and material. The aging effects that required management were identified in the table. A column in the table lists the aging management program that was credited to address the component and aging effect during the period of extended operation. The NRC has reviewed and approved the aging management strategy presented in the LRA, as documented in the SER on license renewal [1]. Table 3.1-1 from the LRA is included in Appendix B as Table B-1.

Duke Energy's commitment to implement MRP-227, Revision 1-A necessitates that the aging management strategy in the original LRA be updated. Catawba utilizes NUREG-1801, Revision 2 as updated by LR-ISG-2011-04 [38] to ensure the aging effects are managed so that the intended function(s) will be maintained consistent with the CLB for the period of extended operation.

The results of the industry research provided by MRP-227, Revision 1-A, summarized in the tables of Appendix C, provide the basis for the required augmented inspections, inspection techniques to permit detection and characterizing of the aging effects (cracks, loss of material, loss of preload, etc.) of interest, prescribed frequency of inspection, and examination acceptance criteria. This supersedes the aging management review performed in the LRA. The information provided in MRP-227-A and MRP-227, Revision 1-A, is rooted in the GALL methodology.

As discussed in Section 4.1.2, core support structures are examined in accordance with Examination Category B-N-3 of ASME Section XI. The inspections credited in the Catawba LRA are based on utilizing these ASME Section XI exams and the augmented inspections derived from MRP-227, Revision 1-A. The MRP-227, Revision 1-A inspections only augment and do not reduce, alter, or otherwise affect the ASME Section XI requirements.

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Westinghouse Non-Proprietary Class 3 5-3 Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.2 GALL REVISION 2 ELEMENT 2: PREVENTIVE ACTIONS GALL Report AMP Element Description MRP-227-A relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]).

Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program, as described in GALL AMP XI.M2, Water Chemistry. [38]

Catawba Unit 1 and Unit 2 Preventive Action The Catawba RVI Program does not prevent degradation due to aging effects; rather, it provides measures for monitoring to detect degradation prior to loss of intended function. Preventative measures to mitigate aging effects such as loss of material and cracking in the primary water system are established and implemented in accordance with the Catawba Chemistry Control Program [14]. A description and applicability to the Catawba RVI Program is provided in the following subsection.

Chemistry Control Program To mitigate aging effects on component surfaces that are exposed to water as process fluid, chemistry programs are used to control water chemistry for impurities (e.g., dissolved oxygen, chloride, fluoride, and sulfate) that accelerate corrosion. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specific limits. The Catawba PWR Chemistry Control Program [14] is based on the current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines [15].

This program is consistent with the corresponding program described in the GALL Report [6] and [38].

The limits of known detrimental contaminants imposed by the Catawba chemistry monitoring program are consistent with the EPRI PWR Primary Water Chemistry Guidelines.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

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Westinghouse Non-Proprietary Class 3 5-4 5.3 GALL REVISION 2 ELEMENT 3: PARAMETERS MONITORED OR INSPECTED GALL Report AMP Element Description The program manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface-breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method, or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components.

For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of fracture toughness on component integrity is indirectly managed by: (1) using visual or volumetric examination techniques to monitor for cracking in the components, and (2) applying applicable reduced fracture toughness properties in the flaw evaluations, in cases where cracking is detected in the components and is extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation. The program uses physical measurements to monitor for any dimensional changes due to void swelling or distortion.

Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, Aging Management Requirements, in MRP-227-A. [38]

Catawba Unit 1 and Unit 2 Parameters Monitored or Inspected The Catawba RVI Program monitors the following aging effects by inspection, in accordance with the guidance of MRP-227, Revision 1-A. Relevant Indications are as defined by Table 5-3 of MRP-227, Revision 1-A (included as Table C-4 of this document), MRP-228, or the associated Existing Program.

1) Cracking Cracking is due to SCC, PWSCC, IASCC, or fatigue/cyclical loading. Cracking is monitored with visual or volumetric examination for evidence of relevant indications. Surface examinations may also be used to supplement visual examinations for detection and sizing of relevant indications.

2) Loss of Material Loss of material is due to wear. Loss of material is monitored with a visual examination for relevant indications, physical measurement, or eddy current testing.

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Westinghouse Non-Proprietary Class 3 5-5

3) Loss of Fracture Toughness Loss of fracture toughness is due to TE or IE. The impact of loss of fracture toughness on component integrity is indirectly managed by monitoring for cracking by using visual or volumetric examination techniques, and by applying applicable reduced fracture toughness properties in flaw evaluations if any detected cracking is determined to be extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation.

4) Changes in Dimension Changes in dimension are due to void swelling or distortion. Changes in dimension are monitored by visual or volumetric examination.

5) Loss of Preload Loss of preload is due to thermal and ISR or irradiation-enhanced creep. Loss of preload is monitored with a visual or volumetric examination for relevant indications that may be indicative of loosening in applicable bolted, fastened, keyed or pinned connections.

The Catawba RVI Program manages the aging effects noted above by the requirements for the Primary Component inspections from Table 4-3 of MRP-227, Revision 1-A (included in Appendix C of this document as Table C-1), the Expansion Component inspections from Table 4-6 of MRP-227, Revision 1-A (included in Appendix C of this document as Table C-2), and the Existing Component inspections from Table 4-9 of MRP-227, Revision 1-A (included in Appendix C of this document as Table C-3). These tables contain requirements to monitor and inspect the RVI through the period of extended operation to address the aging degradation mechanisms.

Appendices B and C of this document provide a detailed listing of the components and subcomponents and the parameters monitored, inspected, and/or tested.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.4 GALL REVISION 2 ELEMENT 4: DETECTION OF AGING EFFECTS GALL Report AMP Element Description The inspection methods are defined and established in Section 4 of MRP-227-A. Standards for implementing the inspection methods are defined and established in MRP-228. In all cases, well-established inspection methods are selected. These methods include volumetric UT examination methods for detecting flaws in bolting and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

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Westinghouse Non-Proprietary Class 3 5-6 Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). VT-3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation enhanced stress relaxation and creep.

The program adopts the guidance in MRP-227-A for defining the Expansion Criteria that need to be applied to the inspection findings of Primary components and for expanding the examinations to include additional Expansion components. RVI component inspections are performed consistent with the inspection frequency and sampling bases for Primary components, Existing Programs components, and Expansion components in MRP-227-A.

In some cases (as defined in MRP-227-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling or distortion.

Inspection coverages for Primary and Expansion RVI components are implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227. [38]

Catawba Unit 1 and Unit 2 Detection of Aging Effects Detection of indications that are required by the ASME Code Section XI ISI Program is well established and field-proven through the application of the Section XI ISI Program [39, 40, and 41]. The Catawba RVI Program implements the augmented inspection requirements of Table 4-3, Table 4-6, and Table 4-9 from MRP-227, Revision 1-A for the Primary, Expansion, and Existing Components, respectively. These are included in Appendix C of this document for reference. These tables include the inspection frequency and sampling bases. For the Expansion Components of MRP-227, Revision 1-A, the Catawba RVI Program implements the expansion requirements of Table 5-3 of MRP-227, Revision 1-A (included in Appendix C of this document as Table C-4).

Inspection can be used to detect physical effects of degradation including cracking, fracture, wear, and distortion. The choice of an inspection technique depends on the nature and extent of the expected damage. The recommendations supporting aging management for the reactor internals, as contained in this report, are built around three (3) basic inspection techniques: (1) visual, (2) ultrasonic, and (3) physical measurement. Inspection standards developed by the industry for the application of these techniques for augmented reactor internals inspections are documented in MRP-228 [53].

VT-3 Examination for General Condition Monitoring One examination method selected for use, which has an extensive history of use for PWR internals, is visual (VT-3) examination. Such visual examinations are relied upon for detection of general degradation WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 5-7 of PWR internals subject to Table IWB-2500-1 Category B-N-3 requirements. VT-3 examinations are conducted to determine the general mechanical and structural condition of components by detecting discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion; and by identifying conditions that could affect operational or functional adequacy of components. This type of examination has been determined to be acceptable for the continued monitoring of many of the internals within the scope of the Catawba RVI Program. When specified, a VT-3 examination is conducted in accordance with the requirements of the MRP-228 Inspection Standard. All examination personnel, equipment, examinations, classification and measurement of indications, and documentation associated with visual examinations will meet the requirements of MRP-228. VT-3 examinations of internals are conducted using remote examination techniques, because of personnel radiation exposure issues.

VT-1 Visual Examinations and EVT-1 Enhanced Visual Examinations Two (2) examination methods selected for use are visual (VT-1) and enhanced visual (EVT-1) examinations. The VT-1 examinations and the EVT-1 examinations were selected where a greater degree of detection capability, as well as sizing capability, is required - over and above the capability inherent in VT-3 examinations to manage the aging effects. Unlike the detection of general degradation conditions by VT-3 examination, VT-1 and EVT-1 examinations are conducted to detect and size discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion. Specifically, VT-1 is used for the detection and sizing of surface discontinuities such as gaps, while EVT-1 is used for the detection of surface breaking flaws.

When specified in these guidelines, a VT-1 examination is conducted in accordance with the requirements of the MRP-228 Inspection Standard. EVT-1 examination is also conducted in accordance with the requirements described for VT-1 examination with additional requirements as specified in the MRP-228 Inspection Standard. All examination personnel, equipment, examinations, classification and measurement of indications, and documentation associated with visual examinations will meet the requirements of MRP-228. VT-1 and EVT-1 examinations of internals are conducted using remote examination techniques, because of personnel radiation exposure issues.

Ultrasonic Testing Another method selected for use is volumetric examination. An ultrasonic examination (UT) was selected where visual or surface examination is unable to detect the effect of the age-related degradation for some PWR internals. For example, IASCC in baffle/former bolts may occur underneath the bolt head and will be undetectable by visual or surface examination unless the bolt is removed and subject to examination over its entire length. When specified in these guidelines, an UT is conducted in accordance with the requirements of the MRP-228 Inspection Standard. All examination personnel, examinations, classification of indications, and documentation associated with UT will meet the requirements of MRP-228. Additionally, technical justifications as described in MRP-228 are required for qualification of ultrasonic examinations. While UT has only been selected for use in these guidelines for detection of aging effects in bolting, UT is also permissible as an alternative or supplement to the specified visual examinations for other configurations such as plates and welds.

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Westinghouse Non-Proprietary Class 3 5-8 Physical Measurement Examination The effects of loss of material caused by wear, the loss of pre-load or clamping force caused by such mechanisms as thermal and irradiation-enhanced stress relaxation, and excessive distortion or deflection caused by void swelling can be managed in some cases by physical measurements. Table C-1 requires guide card wear measurements in accordance with WCAP-17451-P. These guide card wear measurements will be performed in accordance with WCAP-17451-P and the MRP-228 Inspection Standard.

Additionally, technical justifications, as described in MRP-228, are required when determining measurement uncertainty for flaw, degradation, or wear measurements (e.g., guide card wear).

Surface Examination Surface examination, specifically eddy current examination (ET), can supplement either VT-3 or VT-1/EVT-1 examinations. This supplemental examination may thus be used to reject or accept relevant indications. When selected for use, an ET examination is conducted in accordance with the requirements of MRP-228 [53]. The ET examination, as defined in Catawba's response to IEB 88-09 [16], is also considered sufficient to monitor for the applicable aging effect on Bottom-Mounted Instrumentation Thimble Tubes.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.5 GALL REVISION 2 ELEMENT 5: MONITORING AND TRENDING GALL Report AMP Element Description The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227-A and its subsections. Flaw evaluation methods, including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications, are defined in MRP-227-A. The examination and re-examinations that are implemented in accordance with MRP-227-A, together with the criteria specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide for timely detection, reporting, and implementation of corrective actions for the aging effects and mechanisms managed by the program.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in a RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 5-9 inaccessible to the applicable inspection technique and the resulting impact on the intended function(s) of the assembly containing the components. [38]

Catawba Unit 1 and Unit 2 Monitoring and Trending The methods for monitoring, recording, evaluating, and trending the results from the RVI inspections under the Catawba RVI Program are in accordance with MRP-227, Revision 1-A [51], MRP-228 [53],

WCAP-17096-NP-A [11] (as well as interim guidance PWROG-17071-NP [79]), and ASME Section XI

[39, 40, 41, and 43].

Monitoring is accomplished through implementation of the MRP-227, Revision 1-A Primary, Expansion, and Existing inspections (Tables 4-3, 4-6, and 4-9 of MRP-227, Revision 1-A) according to the criteria specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel.

Implementation of these guidelines provides timely detection, reporting, and corrective actions to manage the effects of the age-related degradation mechanisms within the scope of the program. In Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary and Expansion inspections and monitoring recommendations and the Existing programs credited for inspection and aging management. As discussed in MRP-227, Revision 1-A, inspection of the Primary components provides reasonable assurance for demonstrating component capacity to perform the intended functions. Table C-4 in Appendix C identifies the MRP-227, Revision 1-A expansion criteria from the Primary components. If these expansion criteria are met for a component, the associated Expansion component is to be inspected to manage the aging degradation.

Through implementation of the inspection and evaluation requirements of MRP-227, Revision 1-A, the Catawba RVI Program also addresses potential aging effects in the inaccessible portions of components or redundant component populations and the resulting impact on the intended function(s) of the components.

Recording requirements for the MRP-227, Revision 1-A examinations are provided within MRP-228.

Methodologies for evaluation and disposition of relevant indications observed by the inspections are based on Sections 6 and 7.5 of MRP-227, Revision 1-A.

Trending is supported by the implementation requirements documented in Section 7 of MRP-227, Revision 1-A, which includes an NEI-03-08 Needed requirement for data reporting. Consistent reporting of inspection results across all PWR designs will enable the industry to monitor reactor internals degradation on an ongoing industry basis as the period of extended operation moves forward. Reporting of examination results will allow the industry to monitor and trend results and take appropriate preemptive action through updates of the industry guidelines.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

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Westinghouse Non-Proprietary Class 3 5-10 5.6 GALL REVISION 2 ELEMENT 6: ACCEPTANCE CRITERIA GALL Report AMP Element Description Section 5 of MRP-227-A, which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion RVI component examination methods. For RVI components addressed by examinations performed in accordance with the ASME Code,Section XI, the acceptance criteria in IWB-3500 are applicable. For RVI components covered by other Existing Programs, the acceptance criteria are described within the applicable reference document. As applicable, the program establishes acceptance criteria for any physical measurement monitoring methods that are credited for aging management of particular RVI components. [38]

Catawba Unit 1 and Unit 2 Acceptance Criteria The Catawba RVI Program acceptance criteria for the Westinghouse-designed Primary and Expansion component examinations are consistent with MRP-227, Revision 1-A, Section 5. For the Westinghouse-designed Existing Programs components, the acceptance criteria are described within the applicable program documents. The Catawba RVI Program establishes acceptance criteria for the physical measurement monitoring of the control rod guide cards using the criteria from WCAP-17451-P, Revision 2 [60].

Examination acceptance and expansion criteria for the MRP-227, Revision 1-A inspections are provided in Appendix C, Table C-4. The Existing Programs Components in Table C-3 will continue to be examined in accordance with the credited Existing Program requirements. Augmented inspections, as defined by the MRP-227, Revision 1-A requirements included in Appendix C, Table C-1 and Table C-2, that result in relevant indications will be entered into the Corrective Action Program [21] and addressed by appropriate actions that may include: enhanced inspection, repair, replacement, mitigation actions, or analytical evaluations.

Methodologies for evaluation and disposition of relevant indications observed by the inspections are based on Sections 6 and 7.5 of MRP-227, Revision 1-A.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.7 GALL REVISION 2 ELEMENT 7: CORRECTIVE ACTIONS GALL Report AMP Element Description Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. The implementation of the guidance in MRP-227-A, plus the WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 5-11 implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective actions bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Alternative corrective actions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementation. [38]

Catawba Unit 1 and Unit 2 Corrective Action Relevant indications discovered during examinations of the Catawba RVI will be entered into the Corrective Action Program [21] and addressed by appropriate actions that may include: enhanced inspection, repair, replacement, mitigation actions, or analytical evaluations. The Work Management System [37 and 43] is also part of the Corrective Action Program. The Corrective Action Program charges personnel with the responsibility to identify Undesired Conditions (including conditions adverse to quality), requires that conditions adverse to quality be corrected, and, in the case of significant conditions adverse to quality, ensure that the cause of the condition is determined and actions are taken to preclude repetition. The inspection and evaluation guidance of MRP-227, Revision 1-A [51] provide the basis for what relevant conditions adverse to quality must be addressed by the Corrective Action Program. The ASME Section XI Program [39, 40, 41, 43] establishes the Catawba repair and replacement requirements of ASME Code Section XI and will also be credited for this element.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS GALL Report AMP Element Description Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. The implementation of the guidance in MRP-227-A, in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies referenced in this AMP, provides an acceptable level of quality and an acceptable basis for confirming the quality of inspections, flaw evaluations, and corrective actions. [38]

Catawba Unit 1 and Unit 2 Confirmation Process Catawba has an established 10 CFR 50, Appendix B Program [18] that addresses the elements of corrective actions, confirmation process, and administrative controls. Quality Assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. Catawba implements work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [42].

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Westinghouse Non-Proprietary Class 3 5-12 justification for the deviation must be prepared and approved as described in Appendix B of NEI 03-08 and Duke Energy administrative procedures, including notification to the NRC for information only.

The implementation of the guidance in MRP-227, Revision 1-A, in conjunction with the requirements of NEI 03-08 and other guidance documents, reports or methodologies referenced in this document, provides an acceptable level of quality and an acceptable basis for confirming the quality of inspection, flaw evaluation and other elements of aging management of the Catawba RVI.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

5.9 GALL REVISION 2 ELEMENT 9: ADMINISTRATIVE CONTROLS GALL Report AMP Element Description The administrative controls for these types of programs, including their implementing procedures and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable. The evaluation in Section 3.5 of the NRCs SE, Revision 1, on MRP-227 provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a licensee executive. [38]

Catawba Unit 1 and Unit 2 Administrative Controls Catawba has an established 10 CFR 50, Appendix B Program [18] that addresses the elements of corrective actions, confirmation process, and administrative controls. QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B.

Catawba implements work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [42]. A failure to meet a Needed or a Mandatory requirement is a deviation from the guidelines and a written justification for the deviation must be prepared and approved as described in Appendix B to NEI 03-08 and Duke Energy administrative procedures, including notification to the NRC for information only.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

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Westinghouse Non-Proprietary Class 3 5-13 5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE GALL Report AMP Element Description The review and assessment of relevant operating experience for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A.

Consistent with MRP-227-A, the reporting of inspection results and operating experience is treated as a Needed category item under the implementation of NEI 03-08.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience, as discussed in Appendix B of the GALL Report, which is documented in LR-ISG-2011-05. [38]

Catawba Unit 1 and Unit 2 Operating Experience Extensive industry and Catawba OE has been reviewed during the development of the reactor vessel internals AMP and Inspection Plan. The experience reviewed includes NRC Information Notices 84-18, Stress Corrosion Cracking in PWR Systems [19] and 98-11, Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants [20]. Most of the industry OE reviewed has involved cracking of austenitic stainless steel baffle-former bolts, or SCC of high-strength internals bolting. SCC of control rod guide tube support pins has also been reported. Recent OE associated with the clevis bearing stellite wear surfaces, clevis insert bolts, thermal sleeves, and control rod guide tube guide cards has also been reviewed.

Early plant operating experience related to hot functional testing and reactor internals is documented in plant historical records. Inspections performed as part of the ASME Section XI Program [39, 40, 41, and 5] have been conducted as designated by existing commitments and are expected to discover general internals structure degradation. To date, very little degradation has been observed industry-wide.

A review of industry and plant-specific experience with reactor vessel internals reveals that the U.S. industry, including Duke Energy and Catawba, has responded proactively to industry issues relative to reactor internals degradation. An example that demonstrates this proactive response by Duke Energy is the replacement of control rod guide tube support pins at both Catawba Units between 2000 and 2001.

The replacement pins included a material upgrade to Type 316 stainless steel in support of managing aging in the component. Duke Energy will address industry guidance on the Catawba baffle-former bolts, as discussed in Section 4.2.4, on the Catawba core barrel welds, as discussed in Section 4.2.6, on the Catawba clevis bearing stellite wear surface and clevis insert bolts, as discussed in Section 4.2.7, and on the Catawba thermal sleeves, as discussed in Section 4.2.8. Industry related issues with accelerated guide card wear will be managed by Duke Energy as discussed in Section 4.2.3.

A key element of the MRP-227, Revision 1-A guideline is the reporting of age-related degradation of RVI components. Duke Energy, through its participation in PWROG and EPRI/MRP activities, will continue to benefit from the reporting of inspection information and will share its own OE with the industry through the reporting instructions of Section 7 of MRP-227. The collected information from MRP-227, Revision 1-A augmented inspections will benefit the industry in its continued response to RVI aging degradation. Duke Energy will continue to maintain cognizance of industry activities related to WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 5-14 PWR internals inspection and aging management and will address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.

Industry OE is routinely reviewed by Duke Energy using the Institute of Nuclear Power Operations (INPO) OE, the Nuclear Network, and other information sources as directed under the Catawba operating experience procedure [22], for the determination of additional actions and lessons learned.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [6], as updated via LR-ISG-2011-04 [38].

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Westinghouse Non-Proprietary Class 3 6-1 6 MRP-227 SAFETY EVALUATION CONDITIONS AND ACTION ITEMS Revision 0 of MRP-227 was finalized in December 2008 and transmitted to the NRC in January 2009. It was requested that the NRC issue a SE on MRP-227, Revision 0. In June 2011, the NRC issued Revision 0 of an SE for MRP-227, Revision 0 [10]. The NRC issued Revision 1 of the SE in December 2011, which incorporated technical changes required to ensure MRP-227 included all NRC required changes and defined the basis for acceptance of MRP-227. Revision 1 of the SE included eight (8) applicant/licensee-specific action items and seven (7) topical report conditions.

MRP-227, Revision 1 was published in October 2015 and incorporated various updates since the publication of MRP-227-A, as seen in Appendix C of the document. MRP-227, Revision 1 was transmitted to the NRC in December 2015. The NRC issued several RAIs on MRP-227, Revision 1 and later issued an SE on MRP-227, Revision 1 in April 2019 [52]. The accepted version of the I&E guidelines, containing the RAIs and SE, was issued as MRP-227, Revision 1-A in November 2019.

Several MRP-227-A A/LAIs were addressed through generic industry programs, including PWROG-15032-NP [75] and PWROG-14048-P [74] for the CASS components and PWROG-15105-NP [96] for potential cold worked austenitic stainless steel. The rest of the A/LAIs applicable to Catawba were addressed through the revisions to the applicability requirements in MRP-227, Revision 1-A. These applicability requirements are detailed in Section 2.4 and Appendix B of MRP-227, Revision 1-A. The NRC as described within the SE on MRP-227, Revision 1 concluded that the A/LAIs were addressed through the revision or through generic industry programs.

The MRP-227, Revision 1-A, Section 2.4 and Appendix B guideline applicability requirements are addressed in this section. Sections 6.1.1 and 6.1.2, detail how Catawba meets the general assumptions used in the development of MRP-227, Revision 1-A, including the MRP-191 [36] FMECA assumptions and inputs and the guidance on plant-specific fuel design or fuel management provided in MRP-227, Revision 1-A, Appendix B. Section 6.1.3 addresses the items contained in the enclosure to the NRC acceptance letter for MRP-227, Revision 1-A [89].

Additionally, the NRC SE on MRP-227, Revision 1 included a new applicant/license action item for baffle-former bolt degradation. Section 6.2 describes the ways in which this applicant/licensee action item is addressed for Catawba.

6.1 MRP-227, REVISION 1-A GUIDELINE APPLICABILITY The MRP-227, Revision 1-A guidelines are based on a broad set of assumptions about plant operation, which encompass the range of current plant conditions for the domestic fleet of PWRs. The engineering evaluations and assessment and the resultant supporting aging management strategies in MRP-231 and MRP-232 provide the basis for the MRP-227, Revision 1-A guidelines. These evaluations were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter. Subsections 6.1.1 and 6.1.2 describe the guidelines of applicability of MRP-227, Revision 1-A and Catawbas compliance.

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Westinghouse Non-Proprietary Class 3 6-2 The NRC accepted MRP-227, Revision 1-A to the extent delineated within the NRC SE on MRP-227, Revision 1 [52], but requested that utilities who reference MRP-227, Revision 1-A address the items described within [89]. This NRC guidance was transmitted to the industry within MRP 2020-007 [90].

The item within [89] that pertains to Catawba plants is item 5. This item is discussed in detail within subsection 6.1.3 of this document.

6.1.1 General Assumptions The applicability of MRP-227, Revision 1-A to Catawba requires compliance with the following MRP-227, Revision 1-A general assumptions:

• 30 years of operation with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation, as well as the average core power levels and proximity of active fuel to the upper core support plate satisfies limits as described in Appendix B for Westinghouse/CE plants.

Catawba Applicability:

Catawba Unit 1 had approximately 11 years of operation with fresh fuel assemblies at peripheral locations (high leakage core loading pattern) and has been implementing low-leakage core designs since that time [50]. Catawba Unit 2 had approximately 13.6 years of operation with fresh fuel assemblies at peripheral locations (high leakage core loading pattern) and has been implementing low-leakage core designs since that time [50]. The low leakage loading pattern has been applied to all subsequent core designs through current operation and no change to the low-leakage core design philosophy is anticipated for the extended plant operating license [50]. By operating with a high leakage core design for less than 30 years of operation, Catawba has taken a conservative approach.

Appendix B of MRP-227, Revision 1-A provides guidance on plant-specific fuel design and fuel management requirements for the applicability of the MRP-227, Revision 1-A guidelines. A comparison of the Catawba core geometries and operating characteristics to the applicability guidelines for Westinghouse-designed reactors specified in Appendix B was performed for Catawba within [50]. This comparison concludes that Catawba has not utilized atypical fuel design or fuel management that could make the assumptions of MRP-227, Revision 1-A regarding core loading/core design non-representative, including power changes/uprates that have occurred over their operating lifetimes.

Catawba Unit 1 and Unit 2 each operated for less than 30 years with a high-leakage core loading pattern and each operate with fuel design and management that meets the requirements of MRP-227, Revision 1-A, Appendix B. Therefore, Catawba meets the fuel management and neutron fluence applicability requirements of MRP-227, Revision 1-A.

• The power plant has operated for the majority of its lifetime as a base-loaded unit and is currently operating as a base-loaded power plant, in that the unit operates at fixed thermal power levels and does not usually vary power on a calendar or load demand schedule.

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Westinghouse Non-Proprietary Class 3 6-3 Catawba Applicability:

Catawba Unit 1 and Unit 2 currently operate as base-loaded power plants and have operated under base load conditions for the majority of the life of the plant [44]. Therefore, Catawba satisfies this assumption in MRP-227, Revision 1-A regarding operational parameters affecting fatigue.

• No design changes beyond those identified in general industry guidance or recommended by the original vendors.

Catawba Applicability:

Modifications to the Catawba reactor internals include a control rod guide tube support pin replacement at Catawba Unit 1 and Unit 2 [23] a power uprating project at Catawba Unit 1 [30],

and the installation of flux thimble anti-vibration sleeves at Catawba Unit 1. The design has been maintained over the lifetime of the plant as specified by the Original Equipment Manufacturer (OEM) [44]. Therefore, Catawba satisfies this assumption in MRP-227, Revision 1-A.

• The components and material class of each functional component are as listed in the latest revision of MRP-189 or MRP-191, as applicable to the individual plant design.

Catawba Applicability:

The typical Westinghouse PWR internals components are listed within MRP-191, Revision 1, Table 4-4 [36]. The Catawba components were identified in the Catawba AMR [12] and the applicable Catawba RVI component drawings. Once the Catawba specific RVI components were identified, the material of each component was determined from the applicable component drawings [44]. The comparison of typical Westinghouse PWR internals components to the Catawba PWR internals components is described in the following sub list.

a. Confirmation that no additional items were identified by this comparison.

The Catawba components were compared to MRP-191, Revision 1 [36] within [44]. This review identified two components that were not specifically included within MRP-191 Revision 1 [36]: the irradiation specimen holder (spring) and the flux thimble anti-vibration sleeves. The irradiation specimen holder (spring) is a subcomponent of the irradiation specimen plug assembly, which is a subcomponent of the irradiation specimen holder assembly. The irradiation specimen holder (spring) was originally included in MRP-191, Revision 0 [9] and Revision 1 [36] as part of the irradiation specimen plug assembly but the spring piece-part was not specifically listed. The flux thimble anti-vibration sleeves are protective sleeves for the flux thimbles. They were installed within Catawba Unit 1 to reduce the clearance within the lower internals BMI columns and limit the amount of vibratory motion of the flux thimbles. The flux thimble anti-vibration sleeves were not specifically listed alongside the flux thimble tubes or bottom-mounted instrumentation assemblies within MRP-191, Revision 0 [9] or Revision 1 [36]. For MRP-191, Revision 2 [55], both of these components were added to the scope of the expert panel review and screening, categorization, and ranking, as described within Section 6.1.2 of this document. Both the irradiation specimen holder (spring) and the flux thimble anti-vibration sleeves were assigned to safety and economic category A and thus require no additional measures.

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Westinghouse Non-Proprietary Class 3 6-4 Considering these results from MRP-191, Revision 2 [55], the irradiation specimen holder at Catawba Unit 1 and Unit 2 and the flux thimble anti-vibration sleeves at Catawba Unit 1 do not require changes to the MRP-227, Revision 1-A aging management requirements or any further actions.

b. Confirmation that the materials identified for Catawba are consistent with those materials identified in MRP-191, Table 4-4.

All of the materials for Catawba are identical or equivalent to those materials identified in MRP-191, Revision 1, Table 4-4 [36] for Westinghouse-designed plants, except the irradiation specimen plug (spring) and the flux thimble anti-vibration sleeves since these components were not included within MRP-191, Revision 1. Components considered equivalent are those that were fabricated from a different material of the same material class considered in MRP-191, such as a different type of austenitic stainless steel. Examples include components fabricated from Type 316 SS instead of Type 304 SS, Type 304L SS instead of Type 304 SS, or Type 304 SS instead of Type 316 SS. Types 304, 316, and 304L SS fall under the austenitic stainless steel category and there are no differences in the screening criteria for the materials. With no changes to the susceptibility or degradation mechanisms of concern, the FMECA and functionality analysis are still acceptable.

The MRP-191, Revision 2 [55] expert panel review results for the addition of the irradiation specimen plug (spring) and flux thimble anti-vibration sleeves are described in Section 6.1.2.

c. Confirmation that the Catawba internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

The Catawba internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication [44].

• If major plant-specific differences from the inputs to the FMECA process described in MRP-189 and 191 are identified, then plant owners must determine and document the impact, if any, on the aging management strategy described herein.

Catawba Applicability:

Catawba Unit 1 and Unit 2 do not have any major plant-specific differences from the inputs to the FMECA process. The comparison of FMECA inputs to the Catawba RVI is described in the following sub list.

a. Confirmation that the Catawba RVI materials operated at temperatures within the original design basis parameters:

The Catawba Unit 1 reactor coolant system operates between Thot and Tcold, which are not less than approximately 555.3°F for Tcold and not higher than 617.4°F for Thot [44]. The Catawba Unit 2 reactor coolant system operates between Thot and Tcold, which are not less than approximately 558.3°F for Tcold and not higher than 620.8°F for Thot [44]. The design temperature for the reactor vessel is 650°F. The Catawba operating history is within original design basis parameters and therefore consistent with the assumptions used to WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 6-5 develop the MRP-227, Revision 1-A aging management strategy with regard to temperature operational parameters.

b. Confirmation that the Catawba stress values are consistent with the assumptions in MRP-227, Revision 1-A.

Modifications to the Catawba reactor internals made over the lifetime of the plant are those specifically directed by Westinghouse, the OEM [44]. The design has been maintained over the lifetime of the plant as specified by the OEM. Therefore, the Catawba stress values are represented by the assumptions in MRP-191, MRP-232, and MRP-227, Revision 1-A, confirming the applicability of the generic FMECA.

• Plant modifications to PWR internals (e.g., physical changes) made after calendar year 2007 should be reviewed to assess impacts on strategies contained in these guidelines. Plant modifications made or considered after this date should be reviewed to assess impacts on strategies contained in these guidelines.

Catawba Applicability:

Catawba Unit 1 and Unit 2 have not made any plant modifications to the PWR internals within the scope of these guidelines after calendar year 2007. Therefore, Catawba satisfies this assumption in MRP-227, Revision 1-A.

• MRP-227 originally identified that certain CE and Westinghouse PWR internals components which are subject to inspection under existing programs require further plant-specific evaluation to verify the acceptability of the existing programs, or to identify changes to the existing programs which should be implemented to manage the aging of these components for the period of extended operation. If the existing programs are not acceptable, it is necessary to identify and implement changes to the programs to manage aging of applicable components over the period of extended operation.

Catawba Applicability:

Catawba is compliant with the applicable requirements in Table 4-9 of MRP-227, Revision 1-A.

This is detailed in the plant-specific Catawba program documents for ASME Section XI [39],

[40], and [41] and the Catawba flux thimble tube program [26] and [27]. Catawba has a number of programs and activities that support the aging management of the RVI, such as the Catawba Chemistry Control Program [15] and Control Rod Guide Tube Support Pin Replacement Project

[23].

Catawba Unit 1 and Unit 2 replaced the INCONEL Alloy X-750 support pins with 316 SS support pins [23]. As described in MRP-227, Revision 1-A, Section 4.5, the degradation of support pins is not a safety issue but is an asset management concern. The 316 SS support pins are considered a Category A component and are classified as a no additional measures component. Therefore, the 316 SS support pins do not require a plant-specific aging management program and are not included WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 6-6 within Table 4-9 of MRP-227, Revision 1-A. There are no additional inspections required by the supplier per MRP-227, Revision 1-A for the Catawba 316 SS support pins.

Based on the details within this section, Catawba meets the requirements for application of MRP-227, Revision 1-A as a strategy for managing age-related material degradation in reactor internals components.

6.1.2 MRP-191, Revision 2 FMECA The MRP-191, Revision 2 Expert Panel applied the current FMECA approach for Westinghouse designed plants as described within Section 6.2 of MRP-191, Revision 2 [55]. By applying Section 6.2 of [55], the MRP-191, Revision 2 Expert Panel used the same process, definitions, and inputs as the MRP-191, Revision 1 Expert Panel with the addition of separate safety-related consequence and economic consequence rankings and FMECA groups instead of one combined consequence ranking and FMECA group. The MRP-191, Revision 2 Expert Panel also considered 80 years of operation instead of 60 years.

The MRP-191, Revision 2 Expert Panel was composed of experts filling the following FMECA Roles:

• Component design, testing, and repair

• Structural modeling and analysis

• Thermal-hydraulics and systems analysis

• Neutron fluence and radiation analysis

• Materials degradation and failure experience

• Component inspection experience

• Risk assessment

• Inspection requirements

• System function and OE

• Licensing and regulatory interaction The MRP-191, Revision 2 Expert Panel required consistent definitions for some of the key terms and concepts applied throughout assigning likelihood and consequence levels to the RVI components.

Therefore, the following definitions and categories were used throughout the FMECA:

Component Failure: Material degradation of a given component by one or more credible mechanisms identified in the screening evaluation that causes the component to lose its ability to perform its intended design function during normal operation or under accident conditions. Accident conditions include design basis earthquakes or pipe breaks with no credit for the low likelihood of these accidents actually occurring. Cosmetic wear, craze cracking, and plastic deformation (exclusive of springs) were not considered failures.

Failure Likelihood: The probability that component failure(s) will occur during 80 years of operation. The four (4) categories of failure likelihood are defined in Table 6-2 of [55]. Table 6-2 of

[55] is similar to Table 6-2 of [36] except for the consideration of 80 years as opposed to 60 years.

Core Damage: Physical damage to one or more fuel assemblies or other internals components through direct impact with the fuel, flow-jetting, loss of core support/fuel spring hold-down force, loose parts, blockage/diversion of coolant flow, or loss of insertion ability for more than one control rod that would impair the ability to safely shut down the reactor.

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Westinghouse Non-Proprietary Class 3 6-7 Core Damage Likelihood: The conditional probability that component failure(s) results in core damage given that the failure occurs irrespective of the actual failure likelihood. The four (4) safety consequence category descriptions are defined in Table 6-3 of [55]. Table 6-3 of [55] is similar to Table 6-3 of [36] with the exception of the removal of economic impact.

Economic Impact: Economic burden to the utility such as (but not limited to) costs associated with repair or replacement of reactor vessel internals components.

Economic Impact Likelihood: The conditional probability that component failure(s) results in an economic/financial impact given that the failure occurs irrespective of the actual failure likelihood.

The four (4) categories of economic impact likelihood are defined in Table 6-4 of [55]. Table 6-4 of

[55] covers the economic impacts that were removed from Table 6-3 of [36].

Based on failure likelihood and consequence (either safety or economic), the irradiation specimen plug (spring) and flux thimble anti-vibration sleeves were assigned a safety FMECA Group and economic FMECA Group, as seen in Table 6-1 below, which is Table 6-5 in [55]. The FMECA significance groups within Table 6-5 of [55] were modified from those within Table 6-4 of [36] to be more conservative for the low consequence column. This conservative change was implemented by EPRI to place additional weight on components that the panel members expect to experience degradation. This was driven by the impact that actual plant degradation can have on plant operations, as demonstrated by the instances of recent OE at many PWR units in the worldwide fleet.

Table 6-1: Reactor Internals FMECA (Significance) Groups Used by the MRP-191 Revision 2 Expert Panel [55]

Consequence (Either Safety or Economic)

Failure Likelihood Low Medium High High 3 3 3 Medium 2 2 3 Low 1 1 2 None 0 0 0 The MRP-191, Revision 2 Expert Panel discussed the component design, key parameters, and function of the irradiation specimen plug (spring) and flux thimble anti-vibration sleeves, in addition to material information of each component. The MRP-191, Revision 2 Expert Panel then reviewed the geometry, location, and function of each component, identified screened-in degradation mechanisms for each component, and discussed relevant information from previous Westinghouse Pressurized Water Reactor Owners Group (PWROG) sponsored FMECAs and the MRP Issue Management Table [78]. A likelihood of failure was determined for each component based on the component review and a review of degradation and failure experience. A safety consequence and economic consequence ranking was WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 6-8 determined for each component based on a review of the effects and consequences of degradation and failure.

The irradiation specimen plug (spring) was determined to have no screened-in degradation mechanisms since the component experiences negligible stresses. Therefore, the MRP-191, Revision 2 Expert Panel did not evaluate the irradiation specimen plug (spring) for likelihood of failure or safety and economic consequence. The irradiation specimen plug (spring) was assigned a Safety FMECA Group of 0 and an Economic FMECA Group of 0. The irradiation specimen plug (spring) was assigned a Safety Category of A and an Economic Category of A.

The flux thimble anti-vibration sleeves were determined to screen in for IASCC, wear, fatigue, IE, VS, and ISR/IC. The MRP-191, Revision 2 Expert Panel first considered failure of the anti-vibration sleeves and then consequences of failure. If the anti-vibration sleeves were to fail, wear would increase on the flux thimble tubes. This, however, does not compromise safety as the flux thimble tubes can be periodically inspected by eddy current examination. Catawba, for example, inspects by eddy current examination and monitors wall thinning of the flux thimble tubes in order to predict when tubes should be repaired or replaced, as discussed in Section 4.1.3. Leakage due to failure would be detectable at the flux thimble tube seal table. The MRP-191, Revision 2 Expert Panel considered loose parts as a concern; however, the anti-vibration sleeves are not expected to become a loose part based on design.

The MRP-191, Revision 2 Expert Panel assigned a likelihood of failure ranking of Low for the flux thimble anti-vibration sleeves based on a consideration of the screened in degradation mechanisms.

ISR/IC would reduce the anti-vibration sleeves preload, which may allow for more vibration. However, the anti-vibration sleeve would still be locked in place during operation and would still contribute to vibration reduction. IASCC would also reduce preload, but the anti-vibration sleeve is not expected to become a loose part based on design and would still be present during operation to reduce vibration. The effects of IE and VS are considered to be low due to the anti-vibration sleeves proximity to the core. IE and VS are not expected to prevent the anti-vibration sleeves function. The MRP-191, Revision 2 Expert Panel assigned a Safety Consequence of Low and an Economic Consequence of Low for the flux thimble anti-vibration sleeves, since loose parts are not a concern and loss of vibration reduction would not cause core damage. Therefore, the flux thimble anti-vibration sleeves were assigned a Safety FMECA Group of 1 and an Economic FMECA Group of 1. The flux thimble anti-vibration sleeves were assigned a Safety Category of A and an Economic Category of A. Therefore, no change to the MRP-227, Revision 1-A inspection requirements is required for Catawba.

6.1.3 U.S. NRC Acceptance of MRP-227, Revision 1-A The NRC acceptance of MRP-227, Revision 1-A was communicated through a letter to the EPRI MRP

[89] which was then communicated to the industry through MRP 2020-007 [90]. In accordance with [89],

utilities that reference MRP-227, Revision 1-A must address several plant-specific and design-specific items that were identified by the NRC within Table 1 of [89]. The industry responded to these items within MRP 2020-012 [98] and transmitted these responses to the NRC within MRP 2020-013 [99]. The NRC accepted the industry responses and provided a final SE accepting MRP 2020-012 as an addendum or errata of MRP-227, Revision 1-A [100]. Out of the five items listed within Table 1 of [89], only item 5, Confirmation of Changes to Tabular Footnote References for TR Tables 4-1 through 4-6 and 4-8 and 4-9 pertains to Catawba. Item 5 is broken down into specific confirmations for MRP-227, Revision 1-A WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 6-9 Table 4-2, Table 4-3, and Table 4-6. Table 4-2 is for CE designed plants and is not applicable to Catawba.

For Table 4-3, the NRC questioned the note numbering changes for the following components: W1 control rod guide tube assembly guide plates (cards), W3 core barrel assembly upper flange weld, W4 core barrel assembly lower girth weld, and W6 baffle-former assembly baffle-former bolts.

• W1: Component note 7 from Table 4-3 of MRP-227, Revision 1 for component W1 CRGT Guide Plates (Cards) appears as Note 5 in MRP-227, Revision 1-A. The note number changed from 7 to 5 due to the administrative deletion of notes 4 and 5, because neither of those notes were actually referenced in Table 4-3 in MRP-227, Revision 1. The note text was updated in MRP-227, Revision 1-A in accordance with the response to NRC RAI 19 on MRP-227, Revision 1 and the clarifications made on NRC RAI 19 (see Appendix D of MRP-227, Revision 1-A).

• W3 and W4: Component note 8 from Table 4-3 of MRP-227, Revision 1 for components W3 Upper Flange Weld (UFW) and W4 Lower Girth Weld (LGW) appears as Note 6 in MRP-227, Revision 1-A. The note number changed from 8 to 6 due to the administrative deletion of notes 4 and 5, because neither of those notes were actually referenced in Table 4-3 in MRP-227, Revision 1. The examination coverage within this note was updated in MRP-227, Revision 1-A in accordance with the response to NRC RAI 29 (see Appendix D of MRP-227, Revision 1-A).

Component note 10 from Table 4-3 of MRP-227, Revision 1 was removed from component W3 and W4 consistent with the notes included in response to NRC RAI 29. Note 10 from MRP-227, Revision 1 required expansion to the uninspected portion of the W3 and W4 welds during the same outage if significant flaws were found in the Primary inspection sample. The Primary Component coverage for W3 and W4 in Table 4-3 of MRP-227, Revision 1-A is 100% of the accessible weld length of one side of the weld (either side for W3 and the outer diameter for W4) instead of the 25% required in MRP-227, Revision 1. The change in coverage made Note 10 obsolete and unnecessary.

• W6: Component notes 3, 6, and 9 from Table 4-3 of MRP-227, Revision 1 for component W6 Baffle-Former Bolts appear as Notes 3, 4, and 7 in Table 4-3 of MRP-227, Revision 1-A. The note numbers changed from 6 to 4 and from 9 to 7 due to the administrative deletion of notes 4 and 5, because neither of those notes were actually referenced in Table 4-3 in MRP-227, Revision 1. Component notes 8 and 9 were added into MRP-227, Revision 1-A for component W6 in accordance with the response to NRC RAI 8 and the clarification of the response to RAI 8 (see Appendix D of MRP-227, Revision 1-A).

For Table 4-6, the NRC questioned the note numbering changes within Table 4-6 for the following components: W4.2 middle axial welds and W4.3 lower axial welds.

• W4.2 and W4.3: Components W4.2 Middle Axial Welds (MAW) and W4.3 Lower Axial Welds (LAW) in Table 4-6 of MRP-227, Revision 1 included Note 2; however, the table did not include a Note 3, as indicated in [89]. Note 2 was incorrectly referenced for W4.2 and W4.3 in Table 4-6 of MRP-227, Revision 1 because those components already indicated in the Expansion Coverage column that the inspection should be conducted from the OD of the barrel because the ID of the barrel is inaccessible for these two welds.

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Westinghouse Non-Proprietary Class 3 6-10 Component Notes 5 and 6 were added into MRP-227, Revision 1-A for components W4.2 and W4.3 in accordance with the response to NRC RAI 29 (see Appendix D of MRP-227, Revision 1-A).

As described in detail above, the note numbering changes within item 5 of Table 1 [89] have been confirmed to be administrative or consistent with the industry responses to the NRC RAIs on MRP-227, Revision 1.

6.2 MRP-227, REVISION 1 SE APPLICANT/LICENSEE ACTION ITEM 1:

DEGRADATION OF BAFFLE-FORMER BOLTS If the table in MRP 2017-009 indicates that the subsequent inspection interval is not to exceed 6 years (e.g., downflow plants with 3 percent BFBs with indications or clustering, or upflow plants with 5 percent of BFBs with indications or clustering), the plant-specific evaluation to determine a subsequent inspection interval shall be submitted to the NRC for information within one year following the outage in which the degradation was found. Any evaluation to lengthen the determined inspection interval or to exceed the maximum inspection interval recommended in MRP-2017-009 shall be submitted to the NRC for information at least one year prior to the end of the current applicable interval for BFB subsequent examination [52].

Catawba Unit 1 and Unit 2 Compliance:

Catawba Unit 1 and Unit 2 have not yet performed baffle-former bolt inspections. Both Catawba Unit 1 and Unit 2 are categorized as Tier 4 within NSAL-16-1 [46]. As directed in MRP-227, Revision 1-A (in accordance with MRP 2017-009 [47] and MRP-2017-010 [77]), Catawba will perform the baseline UT inspection of baffle-former bolts no later than 35 EFPY.

Catawba will determine a baffle-former bolt re-examination period based on inspection findings. If the inspection findings do not meet the examination acceptance criteria defined in Section 5 of MRP-227, Revision 1-A, the findings will be dispositioned by plant-specific evaluation per the NEI 03-08, Needed Requirement in Section 7.5 of MRP-227, Revision 1-A and Catawba will document and disposition the findings in the Catawba plant corrective action program [21]. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [47] (i.e., 5% of baffle-former bolts with UT or visual indications or clustering as defined in NSAL-16-1 [46]) for either unit is observed, Catawba will follow the interim guidance within MRP 2017-009 [47] to determine the permitted reinspection interval. This plant-specific evaluation will be submitted to the NRC for information within 1 year following the outage in which degradation was found, in accordance with this A/LAI.

MRP 2017-009 provides a limitation on the permitted re-inspection interval to not exceed a maximum of 6 years without further justification through evaluation, if atypical or aggressive baffle-former bolt degradation is discovered. If Catawba chooses to justify a longer reinspection interval by evaluation or to exceed the maximum reinspection interval within MRP 2017-009 by evaluation, Duke Energy will submit this evaluation to the NRC for information at least 1 year prior to the end of the current applicable reinspection interval for baffle-former bolt subsequent examination in accordance with this A/LAI.

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Westinghouse Non-Proprietary Class 3 6-11 Conclusion Catawba will perform baffle-former bolt inspections in accordance with MRP-227, Revision 1-A and the applicable baffle-former bolt interim guidance. If aggressive baffle-former bolt degradation is discovered, as defined in MRP 2017-009, the evaluation used to determine a subsequent baffle-former bolt inspection interval will be submitted to the NRC for information within 1 year following the outage in which degradation was discovered. Any evaluation to lengthen the determined inspection interval or to exceed the maximum inspection interval recommended in MRP-2017-009 will be submitted to the NRC for information at least 1 year prior to the end of the current applicable baffle-former bolt reinspection interval. Therefore, this A/LAI will be fulfilled by Catawba.

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Westinghouse Non-Proprietary Class 3 7-1 7 INSPECTION PLAN AND IMPLEMENTATION SCHEDULE The components identified in Table 7-1 and Table 7-2 cover the Primary Component inspection scope for Catawba Unit 1 and Unit 2, respectively, and are based on Table 4-3 of MRP-227, Revision 1-A as modified by any applicable interim guidance. Additionally, Table C-1 of this document provides the degradation mechanism(s) being managed, the examination coverage, and any associated Expansion Component links for each of the Primary Components.

Table 7-1 and Table 7-2 provide the due dates for the listed Primary Components; these represent the latest opportunity for Catawba to perform the initial examinations and Catawba may elect to perform examinations earlier than the due dates listed. Many of the initial inspection due dates in MRP-227, Revision 1-A are based on the start of the period of extended operation. The period of extended operation begins at midnight on December 6, 2024 for Catawba Unit 1 and at midnight on February 24, 2026 for Catawba Unit 2. Additionally, the Catawba refueling outages corresponding to inspection due dates that are based on EFPY in Table 7-1 and Table 7-2 are conservative estimates based on 18-month fuel cycles and are subject to change. Subsequent exams will be performed in accordance with MRP-227, Revision 1-A as modified by any applicable interim guidance.

The schedule for implementation of aging management requirements for Expansion Components will depend on the findings from the examinations of the Primary components at Catawba. Table C-2 of this document provides the Expansion Components from MRP-227, Revision 1-A, the applicability to Catawba, the degradation mechanism(s) being managed, the examination method/frequency, the examination coverage, and the Primary Component Links.

Table C-4 of this document provides the examination acceptance criteria and expansion criteria for each of the Primary Components. The examinations specified in Table 7-1, Table 7-2, Table C-1, and Table C-2 of this document will be conducted in accordance with the revision of MRP-228 in effect at the time of the examination.

Table C-3 of this document provides the degradation mechanism(s) being managed, the Existing Program being credited (e.g., ASME Section XI, IEB 88-09), the examination method, and the examination coverage for each of the Existing Programs Components from MRP-227, Revision 1-A. The Existing Programs Components will continue to be examined in accordance with the credited Existing Program requirements.

Should a change occur in plant operational practices or should OE result in changes to the projections, appropriate updates will be performed on affected plant documentation in accordance with approved procedures.

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Westinghouse Non-Proprietary Class 3 7-2 Table 7-1: Catawba Unit 1 Primary Component Inspection Plan Catawba Unit 1 Projected RVI Component Due Date EFPY (Note 5) Examination Method Examination Frequency (Refueling Outage) (Note 6)

W1. Control Rod Per the requirements 1R24 28.5 Per the requirements of WCAP-17451-P, including Guide Tube of WCAP-17451-P Fall 2018 subsequent examinations. (Note 1)

Assembly (Note 2)

Guide plate (cards)

W2. Control Rod Enhanced visual No later than 2 refueling outages from the beginning 1R30 37.5 Guide Tube (EVT-1) examination of the license renewal period and subsequent Fall 2027 Assembly examination on a 10-year interval.

Lower flange welds W3. Core Barrel Enhanced visual 1R30 37.5 No later than 2 refueling outages from the beginning Assembly (EVT-1) examination Fall 2027 of the license renewal period and subsequent Upper flange weld examination on a 10-year interval.

(UFW)

W4. Core Barrel Enhanced visual No later than 2 refueling outages from the beginning 1R30 37.5 Assembly (EVT-1) examination of the license renewal period and subsequent Fall 2027 Lower girth weld examination on a 10-year interval.

(LGW)

W6. Baffle-Former Volumetric (UT) 1R28 34.5 Baseline volumetric (UT) examination interval is Assembly examination. Fall 2024 dependent on the plant design. Subsequent Baffle-former bolts examination is dependent on the plant design and the results of the baseline inspection.

(Notes 3 and 4)

W7. Baffle-Former Visual (VT-3) Baseline examination between 20 and 40 EFPY and 1R31 39 Assembly examination subsequent examinations on a 10-year interval. Spring 2029 Assembly WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 Notes:

(1) Due to the timing of the associated NRC reviews of industry documents and issuance of WCAP-17451-P, Revision 2, MRP-227, Revision 1-A states utilities should Use WCAP-17451-P [25], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [61]

and PWROG letter OG-18-46 dated 2/20/2018 [62]. However, Catawba will follow the guidance in WCAP-17541-P, Revision 2 [60] for management of guide card and lower guide tube continuous guidance wear at Catawba Unit 1.

(2) All 53 rodded locations at Catawba Unit 1 were inspected during the Fall 2018 (1R24) outage [34]. Two (2) of the worst worn guide tubes were relocated to unrodded locations.

(3) Catawba Unit 1 is a Tier 4 plant [46]. In accordance with MRP-227, Revision 1-A, Catawba Unit 1 will perform the baseline UT examination at no later than 35 EFPY.

(4) Re-examination periods shall be determined by plant-specific evaluation per the MRP-227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [47] (i.e., 3% of baffle-former bolts with UT or visual indications or clustering* for downflow plants and 5% of baffle-former bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016-021 [48] and MRP 2017-009 [47]) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [51]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

• Clustering is defined per NSAL-16-1, Revision 1 [46] as three (3) or more adjacent defective baffle-former bolts or more than 40% defective baffle-former bolts on the same baffle plate. Untestable bolts should be reviewed on a plant-specific basis consistent with WCAP-17096-NP-A for determination if these should be considered when evaluating clustering.

(5) MRP-227, Revision 1-A Primary components W5.Baffle-former assembly baffle-edge bolts, W8.Alignment and interfacing components internals hold down spring, and W9.Thermal shield assembly thermal shield flexures are not applicable to Catawba Unit 1. See Table C-1 of this document for further detail.

(6) Projected EFPY values were calculated based on the total EFPD of Cycle R24 [34].

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Westinghouse Non-Proprietary Class 3 7-4 Table 7-2: Catawba Unit 2 Primary Component Inspection Plan Catawba Unit 2 Projected RVI Component Due Date EFPY (Note 5) Examination Method Examination Frequency (Refueling Outage) (Note 6)

W1. Control Rod Per the requirements 2R21 25.7 Per the requirements of WCAP-17451-P, including Guide Tube of WCAP-17451-P Fall 2016 subsequent examinations. (Note 1)

Assembly (Note 2)

Guide plate (cards)

W2. Control Rod Enhanced visual No later than 2 refueling outages from the beginning 2R29 37.5 Guide Tube (EVT-1) examination of the license renewal period and subsequent Fall 2028 Assembly examination on a 10-year interval.

Lower flange welds W3. Core Barrel Enhanced visual No later than 2 refueling outages from the beginning 2R29 37.5 Assembly (EVT-1) examination of the license renewal period and subsequent Fall 2028 Upper flange weld examination on a 10-year interval.

(UFW)

W4. Core Barrel Enhanced visual No later than 2 refueling outages from the beginning 2R29 37.5 Assembly (EVT-1) examination of the license renewal period and subsequent Fall 2028 Lower girth weld examination on a 10-year interval.

(LGW)

W6. Baffle-Former Volumetric (UT) Baseline volumetric (UT) examination interval is 2R27 34.5 Assembly examination. dependent on the plant design. Subsequent Fall 2025 Baffle-former bolts examination is dependent on the plant design and the results of the baseline inspection.

(Note 3 and 4)

W7. Baffle-Former Visual (VT-3) Baseline examination between 20 and 40 EFPY and 2R30 39.0 Assembly examination subsequent examinations on a 10-year interval. Spring 2030 Assembly WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 Notes:

(1) Due to the timing of the associated NRC reviews of industry documents and issuance of WCAP-17451-P, Revision 2, MRP-227, Revision 1-A states utilities should Use WCAP-17451-P [25], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [61]

and PWROG letter OG-18-46 dated 2/20/2018 [62]. However, Catawba will follow the guidance in WCAP-17541-P, Revision 2 [60] for management of guide card and lower guide tube continuous guidance wear at Catawba Unit 2.

(2) All 53 rodded locations at Catawba Unit 2 were inspected during the Fall 2016 outage (2R21) [35]. Eight (8) of the worst worn guide tubes were relocated to spare locations within outages 2R21 and 2R22.

(3) Catawba Unit 2 is a Tier 4 plant [46]. In accordance with MRP-227, Revision 1-A, Catawba Unit 2 will perform the baseline UT examination at no later than 35 EFPY.

(4) Re-examination periods shall be determined by plant-specific evaluation per the MRP-227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [47] (i.e., 3% of baffle-former bolts with UT or visual indications or clustering* for downflow plants and 5% of baffle-former bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016-021 [48] and MRP 2017-009 [47]) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [51]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

• Clustering is defined per NSAL-16-1, Revision 1 [46] as three (3) or more adjacent defective baffle-former bolts or more than 40% defective baffle-former bolts on the same baffle plate. Untestable bolts should be reviewed on a plant-specific basis consistent with WCAP-17096-NP-A for determination if these should be considered when evaluating clustering.

(5) MRP-227, Revision 1-A Primary components W5.Baffle-former assembly baffle-edge bolts, W8.Alignment and interfacing components internals hold down spring, and W9.Thermal shield assembly thermal shield flexures are not applicable to Catawba Unit 2. See Table C-1 of this document for further detail.

(6) Projected EFPY values were calculated based on the total EFPD of Cycle R23 [35].

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Westinghouse Non-Proprietary Class 3 8-1 8

SUMMARY

AND CONCLUSIONS This report documents and provides a description of the Catawba RVI AMP and Inspection Plan and how it relates to the Catawba RVI Program for management of aging effects through the period of extended operation. This Catawba RVI AMP and Inspection Plan is based on MRP-227, Revision 1-A. Section 5 of this document demonstrates that the Catawba RVI Program will meet the intent of the ten (10) AMP elements described in NUREG-1801, Revision 2, Chapter XI, AMP XI.M16A [6], as modified by LR-ISG-2011-04 [38]. Section 6 of this document confirms the applicability of MRP-227, Revision 1-A to Catawba and addresses the applicant/licensee plant-specific action item from MRP-227, Revision 1-A.

The Catawba RVI Program will include this AMP and Inspection Plan and will demonstrate that the program adequately manages the effects of aging for RVI components and establishes the basis for providing reasonable assurance that the RVI components will remain functional through the period of extended operation.

Duke Energy is revising its commitments for RVI Inspections from those that currently exist in the Catawba UFSAR to the inspection guidelines provided in MRP-227, as approved by the NRC. This Catawba RVI AMP and Inspection Plan, as documented herein, is based on MRP-227, Revision 1-A.

Once the Catawba RVI AMP and Inspection Plan is approved by the NRC, the Catawba UFSAR will be updated as required.

WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

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Westinghouse Non-Proprietary Class 3 9-2

14. Duke Administrative Procedure, AD-CP-PWR-0001, Rev. 1, Primary Water Chemistry Program, May 16, 2019.

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22. Duke Energy Administrative Procedure, AD-PI-ALL-0400, Rev. 7, Operating Experience Program, May 1, 2018.

23. Westinghouse Report, WCAP-15252, Rev. 2, Duke Energy Corporation, McGuire and Catawba Units 1 & 2 Replacement CW 316 Support Pin Design Qualification Report, October 2018.

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Westinghouse Domestic Fleet Operational Projections, October 2013.

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28. Duke Energy Letter, RA-14-0005, Letter of Intent to Submit Plant Specific Reactor Vessel Internals Inspection Plans, March 19, 2014 (ML14079A523).

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Westinghouse Non-Proprietary Class 3 9-3

29. Duke Energy License Amendment Request, Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate, June 23, 2014 (ML14176A109).

30. U.S. Nuclear Regulatory Commission Letter, Catawba Nuclear Station, Units 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (CAC Nos.

MF4526 and MF4527), April 29, 2016 (ML16081A333).

31. Duke Energy Administrative Procedure, AD-EG-ALL-1911, Rev. 3, Reactor Internals Program Implementation, June 11, 2019.

32. Duke Energy Letter, Letter of Intent to adopt Materials Reliability Program 227, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 16, 2010 (ML16084A748).

33. Westinghouse Technical Bulletin, TB-14-5, Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation, August 25, 2014.

34. Westinghouse Report, WCAP-18432-P, Rev. 0, Catawba Unit 1 Upper Internals 17x17A Guide Tube - Guide Card Wear Evaluation, February 2019.

35. Westinghouse Report, WCAP-18198-P, Rev. 2, Catawba Unit 2 Upper Internals 17x17A Guide Tube - Guide Card Wear Evaluation, February 2019.

36. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191, Revision 1). EPRI, Palo Alto, CA: 2016. 3002007960.

37. Duke Energy Administrative Procedure, AD-WC-ALL-0210, Rev.7, Work Request Initiation, Screening, Prioritization and Classification, April 26, 2017.

38. LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components For Pressurized Water Reactors, June 3, 2013 (ML12270A436).

39. Duke Energy Administrative Procedure, AD-EG-ALL-1701, Rev.2, ASME Section XI Plan Development, March 14, 2019.

40. Duke Energy Administrative Procedure, AD-EG-ALL-1702, Rev. 6, ASME Section XI Inservice Inspection Program Administration, March 14, 2019.

41. Duke Energy Administrative Procedure, AD-EG-ALL-1704, Rev 2, Augmented Inservice Inspection Program Administration, March 14, 2019.

42. Duke Energy Administrative Procedure, AD-EG-ALL-1912, Rev. 0, Materials Degradation Management Program Implementation, January 14, 2015.

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Westinghouse Non-Proprietary Class 3 9-4

43. Duke Energy Administrative Procedure, AD-EG-ALL-1703, Rev. 2, ASME Section XI Repair/Replacement Program Administration.

44. Westinghouse Calculation Note, CN-RIDA-13-60, Rev. 2, Catawba Units 1 and 2 Reactor Internals MRP-227, Revision 1-A Guideline Applicability and Licensee Action Item 1, June 16, 2020.

45. Westinghouse Letter, NS-TMA-2214, Control Rod Guide Tube Support Pin Cracking, March 14, 1980.

46. Westinghouse Nuclear Safety Advisory Letter, NSAL-16-1, Rev. 1, Baffle-Former Bolts, August 1, 2016.

47. Materials Reliability Program, MRP 2017-009, Transmittal of NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev. 1, March 15, 2017.

48. Materials Reliability Program, MRP 2016-021, Transmittal of NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as Defined in Westinghouse NSAL 16-01, July 25, 2016.

49. Westinghouse Nuclear Safety Advisory Letter, NSAL-17-1, Guide Tube Guide Card Wear Attributed to Ion Nitride Rod Cluster Control Assembly, January 16, 2017.

50. Westinghouse Letter, LTR-REA-19-131, Rev. 0, Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Requirements for Catawba Unit 1 and Unit 2, January 28, 2020.

51. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A). EPRI, Palo Alto, CA: 2019. 3002017168.

52. U.S. Nuclear Regulatory Commission Safety Evaluation, Final Safety Evaluation for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluations Guideline (CAC NO.MF7223; EPID L-2016-TOP-0001), April 25, 2019.

53. Materials Reliability Program: Inspection Standard for Pressurized Water Reactor Internals -

2018 Update (MRP-228, Rev. 3). EPRI, Palo Alto, CA: 2018. 3002010399.

54. Westinghouse Report, WCAP-12866, Rev. 0, Bottom Mounted Instrumentation Flux thimble Wear, January 1991.

55. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191, Revision 2). EPRI, Palo Alto, CA: 2018. 3002013220.

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Westinghouse Non-Proprietary Class 3 9-5

56. Duke Energy Letter, CNS-14-131, License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate Response to NRC Requests for Additional Information (RAIs) (TAC Nos. MF4526 and MF4527), December 15, 2014 (ML14353A027).

57. Duke Energy Letter, CNS-15-007, License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate Response to NRC Requests for Additional Information (RAIs) (TAC Nos. MF4526 and MF4527), January 22, 2015 (ML15029A417).

58. Duke Energy Letter, CNS-15-024, License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate Response to NRC Requests for Additional Information (RAIs) (TAC Nos. MF4526 and MF4527), April 23, 2015 (ML15117A011).

59. Duke Energy Letter, CNS-15-078, License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate Response to NRC Requests for Additional Information (RAIs) (TAC Nos. MF4526 and MF4527), November 16, 2015 (ML15324A083).

60. Westinghouse Report, WCAP-17451-P, Rev. 2, Reactor Internals Guide Tube Wear -

Westinghouse Domestic Fleet Operational Projections, November 2018.

61. Letter from David Czufin and Brian Burgos to the PMMP Executive Committee Members and MRP Integration Committee Members,

Subject:

Transmittal of NEI-03-08 Needed Interim Guidance to Address Accelerated Guide Card Wear Operating Experience (OE) Discussed in NSAL-17-1, MRP Letter 2018-007, March 7, 2018.

62. Pressurized Water Reactor Owners Group Letter, OG-18-46, Transmittal of Approved Needed Interim Guidance for Addressing Accelerated Guide Card Wear Issue Described in NSAL-17-1, (LTR-RIDA-17-270, Revision 0), PA-MSC-1471, February 20, 2019.

63. Pressurized Water Reactor Owners Group Letter, OG-18-276, Guidance for Addressing Accelerated Guide Card Wear Issue Described in WCAP-17451-P Revision 2 Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet - Operational Projections, PA-MSC-1471, November 9, 2018

64. Pressurized Water Reactor Owners Group Letter, OG-19-197, Submittal of WCAP-17451-P Revision 2 Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet - Operational Projections to the NRC for Information Only (PA-MSC-1471), February 20, 2018

65. Letter from Mike Hoehn II and Brian Burgos to the MRP Members,

Subject:

PWR Owners Group Transmittal of Interim Guidance for Addressing Accelerated Guide Card Wear Issue Described in NSAL-17-1, MRP Letter 2018-008, April 5, 2018.

66. Westinghouse Technical Bulletin, TB-07-2, Rev. 3, Reactor Vessel Head Adapter Thermal Sleeve Wear, December 7, 2015.

67. Westinghouse Nuclear Safety Advisory Letter, NSAL-18-1, Thermal Sleeve Flange Wear Leads to Stuck Control Rod, July 9, 2018.

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Westinghouse Non-Proprietary Class 3 9-6

68. Pressurized Water Reactor Owners Report, PWROG-16003-P, Rev. 2, Evaluation of Potential Thermal Sleeve Flange Wear, May 7, 2019.

69. Pressurized Water Reactor Owners Group Letter, OG-19-101, Transmittal of Final PWROG Report PWROG-16003-P, Revision 2, Evaluation of Potential Thermal Sleeve Flange Wear (PA-MSC-1654R1), May 13, 2019.

70. Westinghouse Letter, LTR-RIDA-18-100, Rev. 0, Evaluation of Catawba Unit 2 Thermal Sleeve Flange Wear Inspection Results, March 27, 2018.

71. Westinghouse Letter, LTR-RIDA-18-323, Rev. 0, Evaluation of Thermal Sleeve Flange Wear Condition and Recommendations for Penetration 18 for Catawba Unit 1 (DCP) Fall 2018 Inspection, November 30, 2018.

72. Letter from David Czufin and Brian Burgos to the PMMP Members and MRP RIC Members,

Subject:

NEI-03-08 Needed Inspection Guidance for PWR CRDM Thermal Sleeve Wear, MRP Letter 2018-027, August 31, 2018.

73. Westinghouse Letter, LTR-NRC-18-34, Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21, May 23, 2018.

74. Pressurized Water Reactor Owners Group Report, PWROG-14048-P, Rev. 2, Functionality Analysis: Lower Support Columns, August 2017.

75. Pressurized Water Reactor Owners Group Report, PWROG-15032-NP, Rev. 0, PA-MSC-1288 Statistical Assessment of PWR RV Internals CASS Materials, November 2015.

76. Letter from David Czufin and Brian Burgos to the PMMP Members and MRP Integration Committee Members,

Subject:

Transmittal of NEI-03-08 Needed Interim Guidance Regarding MRP-227-A and MRP-227, Revision 1 Baffle-Former Bolt Expansion Inspection Requirements for PWR Plants, MRP Letter 2018-002, January 17, 2018.

77. Letter to MRP Integration Committee, MRP Assessment TAC, and MRP Inspection TAC,

Subject:

Summary White Paper of the Baffle-Former Bolt Prediction Results Provided by Structural Integrity Associates, AREVA, and Westinghouse, MRP Letter 2017-010, March 17, 2017.

78. Materials Reliability Program: Pressurized Water Reactor Issue Management Tables - Revision 3 (MRP-205). EPRI, Palo Alto, CA: 2013. 3002000634.

79. Pressurized Water Reactor Owners Group Report, PWROG-17071-NP, Rev. 0, WCAP-17096-NP-A Interim Guidance, March 2018.

80. Pressurized Water Reactor Owners Group Letter, OG-18-61, Transmittal of Final Report, PWROG-17071-NP, Revision 0, WCAP-17096-NP-A Interim Guidance (PA-MSC-1567),

March 15, 2018.

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Westinghouse Non-Proprietary Class 3 9-7

81. Westinghouse Report, WCAP-17096-NP, Rev. 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, July 2019.

82. Pressurized Water Reactor Owners Group Letter, OG-19-164, Transmittal of WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, PA-MSC-1567, July 31, 2019.

83. Letter from David Czufin and Brian Burgos to PWR Plant Site Vice Presidents,

Subject:

Notification of Recent PWR Core Barrel Operating Experience and Recommended Plant Actions, MRP Letter 2018-028, August 9, 2018.

84. Letter from Mike Hoehn and Brian Burgos to PMMP Members and MRP Research Integration Committee Members,

Subject:

Transmittal of NEI 03-08 Good Practice Interim Guidance Regarding MRP-227-A and MRP-227, Revision 1 PWR Core Barrel and Core Support Barrel Inspection Requirements, MRP Letter 2019-009, July 17, 2019.

85. Pressurized Water Reactor Owners Group Report, PWROG-16003-P, Rev. 1, Evaluation of Potential Thermal Sleeve Flange Wear, August 4, 2017.

86. Westinghouse Technical Bulletin, TB-16-4, Fuel Alignment Pin Malcomized Surface Degradation, August 15, 2016.

87. Westinghouse Field Service Report, DPC-00-077, Catawba Unit 1 Guide Tube Support Pin Replacement Project, November 13, 2000.

88. Westinghouse Field Service Report, DPC-01-061, Catawba Unit 2 Guide Tube Support Pin Replacement Project, October 12, 2001.

89. U. S. Nuclear Regulatory Commission Letter, U.S. Nuclear Regulatory Commission Verification Letter for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline (L-2019-TOP-0053), February 19, 2020 (ML20006D152).

90. Letter from Mike Hoehn and Brian Burgos to Pressurized Water Reactor Materials Management Program Executive Committee MRP Research Integration Committee,

Subject:

NRC Acceptance of MRP-227 Revision 1-A Topical Report entitled Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline for PWR Licensee Referencing, MRP Letter 2020-007, March 18, 2020.

91. Westinghouse Nuclear Safety Advisory Letter, NSAL-20-1, Reactor Vessel Head Control Rod Drive Mechanism Penetration Thermal Sleeve Cross-Sectional Failure, February 14, 2020.

92. Westinghouse Letter, LTR-NRC-19-79, Notification of the Potential Existence of Defect Pursuant to 10 CFR Part 21, December 12, 2019.

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Westinghouse Non-Proprietary Class 3 9-8

93. Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230, Revision 2). EPRI, Palo Alto, CA: 2012.

1021026

94. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1). EPRI, Palo Alto, CA: 2017.

3002010268.

95. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internal Components (MRP-232, Revision 1). EPRI, Palo Alto, CA: 2012.

1021029.

96. Pressurized Water Reactor Owners Group Report, PWROG-15105-NP, Rev. 0, PA-MSC-1288 PWR RV Internals Cold-Work Assessment, April 2016.

97. Pressurized Water Reactor Owners Group Letter, OG-20-113, NEI 03-08 Needed and Good Practice Guidance: Thermal Sleeve Cross-Sectional Failure-Westinghouse Nuclear Safety Advisory Letter NSAL-20-1, April 13, 2020.

98. Letter from Mike Hoehn and Brian Burgos to the U.S. Nuclear Regulatory Commission, Responses to Nuclear Regulatory Commission Staff Comments on MRP-227, Revision 1-A, MRP Letter 2020-012, May 4, 2020.

99. Letter from Chris Koehler and Brian Burgos to the U.S. Nuclear Regulatory Commission,

Subject:

Industry response to draft safety evaluation of EPRI report MRP-227-Revision 1-A, MRP Letter 2020-013, May 22, 2020 (ML20146A002).

100. U.S. Nuclear Regulatory Commission Safety Evaluation, Final SE MRP-227, Rev 1-A Supplemental Information, June 17, 2020 (ML20141L313).

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A ILLUSTRATIONS Figure A-1: Illustration of Typical Westinghouse Internals WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-2 Figure A-2: Typical Westinghouse Control Rod Guide Card WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-3 Figure A-3: Typical Lower Section of Control Rod Guide Tube Assembly WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-4 Figure A-4: Major Core Barrel Welds WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-5 er rm Fo ts e

affl Bol B

Baffle Edge Bolts Figure A-5: Bolting Systems Used in Westinghouse Core Baffles (Note: Baffle-Edge Bolts are not applicable to Catawba Unit 1 and Unit 2)

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Westinghouse Non-Proprietary Class 3 A-6 Figure A-6: Core Baffle/Barrel Structure WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-7 Figure A-7: Bolting in a Typical Westinghouse Baffle-Former Structure (Note: Baffle-Edge Bolts and Corner Edge Bracket Baffle to Former Bolts are not applicable to Catawba Unit 1 and Unit 2)

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Westinghouse Non-Proprietary Class 3 A-8 Figure A-8: Lower Core Support Structure WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-9 Figure A-9: Lower Core Support Structure - Core Support Forging Cross-Section WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-10 Figure A-10: Typical Core Support Column WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-11 Figure A-11: Typical Bottom-Mounted Instrumentation (BMI) Column Design WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 A-12 Figure A-12: Typical Westinghouse-Design Upper Internals Assembly WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B CATAWBA UNIT 1 AND UNIT 2 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLES The content and numerical identifiers in Table B-1 are extracted from Table 3.1-1, Aging Management Review Results - Reactor Coolant System, of the Catawba LRA. Only those items applicable to reactor vessel internals (according to the LRA) were imported into Table B-1 from the LRA.

Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Reactor Vessel Internals Upper Core Support Structure Upper Support Assembly 1, 2, 3, 4 Stainless Steel Borated Water Cracking Chemistry Control (Forging, Plates, Weld) Program Loss of Material Inservice Inspection Plan Upper Support Column 1, 2, 4 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Upper Support Column 1, 2, 4 CASS Borated Water Cracking Chemistry Control (Base, Conduit Support, Program Thermocouple stop (U1))

Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 B-2 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Upper Support Column 1, 2, 4 Stainless Steel Borated Water Cracking Chemistry Control Bolts Program Loss of Material Inservice Inspection Plan Loss of Preload Reactor Vessel Internals (bolting) Inspection Upper Core Plate 1, 2, 3, 4 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Dimensional Reactor Vessel Internals Changes Inspection Reduction in Fracture Toughness Upper Core Plate 1 Stainless Steel Borated Water Cracking Chemistry Control Alignment Pins Program Loss of Material Inservice Inspection Plan Dimensional Reactor Vessel Internals Changes Inspection WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 B-3 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Fuel Alignment Pins 1 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Dimensional Reactor Vessel Internals Changes Inspection Loss of Preload Hold-down Spring 1 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Loss of Preload Thermocouple Column and 4 Stainless Steel Borated Water Cracking Chemistry Control Crossrun Assemblies Program Loss of Material Inservice Inspection Plan 17x17 and 15x15 Guide 2 (17x17 only), 3 Stainless Steel Borated Water Cracking Chemistry Control Tube Assembly Program Loss of Material Inservice Inspection Plan WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 B-4 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities 15x15 and 17x17 Guide 3 CASS Borated Water Cracking Chemistry Control Tube Assembly Program Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection UHI Flow Columns 3 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan UHI Flow Columns (Base) 3 CASS Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-5 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Lower Core Support Structure Core Barrel Flange 1, 3, 4, 5, 6 Stainless Steel Borated Water Cracking Chemistry Control Core Barrel Outlet Nozzles Program Neutron Panels Irradiation Specimen Loss of Material Inservice Inspection Plan Holder Fasteners Reduction in Reactor Vessel Internals Fracture Toughness Inspection Dimensional Changes Loss of Preload (bolting)

Irradiation Specimen 5 Nickel Based Borated Water Loss of Material Alloy 600 Aging Holder (spring) Alloy Management Review (Note 3)

Cracking Inservice Inspection Plan Chemistry Control Program WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 B-6 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Baffle and Former Plates 1, 3, 6 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection Dimensional Changes Baffle Bolts (baffle to 1, 3 Stainless Steel Borated Water Cracking Chemistry Control baffle, baffle to former) Program Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection Dimensional Changes Loss of Preload (bolting)

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• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-7 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Lower Core Plate 1, 3, 4, 5 Stainless Steel Borated Water Cracking Chemistry Control Fuel Alignment Pins Program Lower Support Column Bolts Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection Dimensional Changes Loss of Preload (bolting)

Lower Support Plate 1, 3, 4, 5, 6 Stainless Steel Borated Water Cracking Chemistry Control (forging) Program Lower Core Support Columns Loss of Material Inservice Inspection Plan Radial Keys and Fasteners 1 Stainless Steel Borated Water Cracking Chemistry Control Program Loss of Material Inservice Inspection Plan WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-8 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Clevis Inserts and Fasteners 1 Nickel-Based Borated Water Cracking Alloy 600 Aging Alloy Management Review Loss of Material Chemistry Control Program Inservice Inspection Plan Bottom-Mounted Instrumentation Bottom-Mounted 3, 4, 5 Stainless Steel Borated Water Cracking Chemistry Control Instrumentation Program (Plates, forgings, welds, energy absorber, fasteners) Loss of Material Inservice Inspection Plan Bottom-Mounted 4 CASS Borated Water Cracking Chemistry Control Instrumentation Program (upper end, cruciform)

Loss of Material Inservice Inspection Plan Reduction in Reactor Vessel Internals Fracture Toughness Inspection Thimble Assembly PB Stainless Steel Borated Water Loss of Material Chemistry Control Program Cracking Bottom Mounted Thimble Tube Instrumentation Inspection Program WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-9 Table B-1: Aging Management Review Results - Reactor Coolant System (Reactor Vessel Internals Components)

Component Function Material Aging Management Component Type (Note 1) (Note 2) Environment Aging Effect Programs and Activities Table B-1 Notes:

(1) Reactor Vessel Internals Functions:

1. Provide support and orientation of the reactor core (i.e., the fuel assemblies).

2. Provide support, orientation, guidance and protection of the control rod assemblies.

3. Provide a passageway for the distribution of the reactor coolant flow to the reactor core.

4. Provide a passageway for the support, guidance, and protection for the in-core instrumentation.

5. Provide secondary core support for limiting the downward displacement of the core support structure in the event of a postulated failure of the core barrel.

6. Provide neutron shielding to the reactor vessel and provide support for the vessel material test specimens.

(2) Material:

Alloy Steel refers to low alloy steel.

Nickel Based Alloys include Inconel and Alloys 600 and 690.

CASS Cast Austenitic Stainless Steel.

(3) Component

Description:

This item specifically refers to the Access Plug Assembly Spring. The Irradiation Specimen Holder (spring) is a subcomponent of the Access Plug Assembly which is a subcomponent of the Irradiation Specimen Holder Assembly. The Irradiation Specimen Holder (spring) is labled as the Irradiation Specimen Plug (spring) within MRP-191, Revision 2 [55].

(4) Reactor Vessel and CRDM Pressure Boundary Components:

PB: Maintain mechanical pressure boundary integrity so that sufficient flow and/or sufficient pressure are delivered, effect Containment isolation for fission product retention, or prevent physical interaction with safety-related equipment.

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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C MRP-227 AUGMENTED INSPECTIONS Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage W1.Control Rod Guide All plants Loss of Material None Per the requirements of Examination coverage per Tube Assembly (See WEC (Wear) WCAP-17451-P, including the requirements of Guide plates (cards) NSAL-17-1) subsequent examinations. WCAP-17451-P, Revision (Note 5) 1. (Note 5)

See Figure A-2.

W2.Control Rod Guide All plants Cracking (SCC, W2.1.Remaining Enhanced visual (EVT-1) 100% of outer (accessible)

Tube Assembly Fatigue) CRGT assembly examination to determine the CRGT lower flange weld Lower flange welds Aging lower flange presence of crack-like surfaces and 0.25-inch of Management (IE welds surface flaws in flange welds the adjacent base metal on and TE) no later than 2 refueling the individual periphery outages from the beginning CRGT assemblies.

W2.2.BMI of the license renewal period (Note 2) column bodies and subsequent examination on a 10-year interval.

See Figure A-3.

W3.Core Barrel All plants Cracking (SCC) W3.1.Upper girth Enhanced visual (EVT-1) 100% of the accessible Assembly weld (UGW), examination, no later than 2 weld length of one side of Upper flange weld W3.3.lower refueling outages from the the UFW and 3/4 of (UFW) flange weld beginning of the license adjacent base metal shall (LFW), renewal period and be examined.

W3.2.Upper axial subsequent examination on a (Note 6) welds (UAW), 10-year interval.

W3.4.Lower support forging See Figure A-4.

or casting WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-2 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage W4.Core Barrel All plants Cracking W4.1.Upper core Enhanced visual (EVT-1) 100% of the accessible Assembly (SCC, IASCC), plate, examination, no later than 2 weld length of the OD of Lower girth weld Aging W4.4.Lower refueling outages from the the LGW and 3/4 of (LGW) Management (IE) support column beginning of the license adjacent base metal shall bodies (cast, non- renewal period and be examined.

cast), subsequent examination on a (Note 6)

W4.2.Middle 10-year interval.

axial welds (MAW), See Figure A-4.

W4.3.Lower axial welds (LAW)

W5.Baffle-Former All plants Cracking None Visual (VT-3) examination, Bolts and locking devices Assembly with baffle- (IASCC, Fatigue) with baseline examination on high-fluence seams.

Baffle-edge bolts edge bolts that results in: between 20 and 40 EFPY and 100% of components

• Lost or broken subsequent examinations on accessible from core side.

locking devices a 10-year interval.

Not applicable to

• Failed or See Figures A-5, A-6, and Catawba Unit missing bolts A-7.

1 and Unit 2

• Protrusion of bolt heads Aging Management (IE and ISR)

(Note 4)

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Westinghouse Non-Proprietary Class 3 C-3 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage W6.Baffle-Former All plants Cracking W6.2.Lower Baseline volumetric (UT) 100% of accessible bolts.

Assembly (See WEC (IASCC, Fatigue) support column examination interval is (Note 3)

Baffle-former bolts NSAL-16-1) Aging bolts, dependent on the plant Management W6.1.Barrel- design (Note 8). Subsequent (Note 7) examination is dependent on See Figures A-5, A-6, and (IE and ISR) former bolts the plant design and the A-7.

(Note 4) results of the baseline inspection (Note 9).

W7.Baffle-Former All plants Distortion None Visual (VT-3) examination to Core side surface:

Assembly (Void Swelling), check for evidence of

• High fluence baffle Assembly or Cracking distortion, with baseline joints (Includes: Baffle plates, (IASCC) that examination between 20 and results in: 40 EFPY and subsequent

• Top and bottom edge of baffle edge bolts, corner baffle plates bolts, and indirect

• Abnormal examinations on a 10-year effects of void swelling interaction with interval.

• Bolts and locking in former plates) fuel assemblies devices

• Gaps between plates See Figure A-6.

• Vertical displacement of baffle plates

• Broken or damaged edge bolts WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-4 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage W8.Alignment and All plants Distortion (Loss None Direct measurement of Measurements should be Interfacing with 304 of Load due to spring height within 3 cycles taken at several points Components stainless steel Stress of the beginning of (before or around the circumference Internals hold-down hold-down Relaxation) after) the license renewal of the spring, with a spring springs period. If the first set of statistically adequate NOTE: measurements is not number of measurements sufficient to assess remaining at each point to minimize The Catawba life, additional spring height uncertainty.

Unit 1 and measurements will be Unit 2 hold-required.

down spring is 403 SS; therefore, this component is not applicable to Catawba Unit 1 and Unit 2.

W9.Thermal Shield All plants Cracking None Visual (VT-3) no later than 2 100% of accessible Assembly with thermal (Fatigue) or Loss refueling outages from the surfaces of 100% of Thermal shield flexures shields (See of Material beginning of the license thermal shield flexures.

WEC (Wear) that renewal period. Subsequent (Note 10)

TB-19-5) results in thermal examinations on a 10-year shield flexures interval.

excessive wear, Not fracture, or applicable to complete Catawba Unit separation 1 and Unit 2 WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-5 Table C-1 Notes:

(1) Examination acceptance criteria and expansion criteria for the Westinghouse components are listed in Table C-4.

(2) A minimum of 75% of the total identified sample population must be examined.

(3) A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table C-4:, must be examined for inspection credit.

(4) Void swelling effects on this component are managed through management of void swelling on the entire baffle-former assembly.

(5) In WCAP-17451-P the baseline examination schedule has been adjusted for various CRGT designs, the extent of individual CRGT examination modified, and flexible subsequent examination regimens correlating to initial baseline sample size, accuracy of wear estimation and examination results. Initial inspection prior to the license renewal period may be required. Use WCAP-17451-P [25], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [61] and PWROG letter OG-18-46 dated 2/20/2018 [62]. Note, Catawba Unit 1 and Unit 2 will follow the inspection and evaluation guidance within WCAP-17451-P, Revision 2, which includes the modified requirements provided in MRP 2018-007 and OG-18-46.

(6) Examination coverage requires a minimum of 50% of the length of either the ID or the OD of the weld being examined.

(7) Baffle-former bolt inspection includes inspection of the corner plate bolts when applicable.

(8) In accordance with MRP 2017-009 [47] and MRP 2017-010 [77], Tier 1 plants are to perform the baseline UT examination by 20 EFPY or during the next refueling outage after March 1, 2016. Per MRP 2017-009 [47], Tier 2 plants are to perform the baseline UT examination at no later than 30 EFPY (initial Tier 2 plant baseline UT exams performed prior to 1/1/2018 are acceptable). All other remaining plants are to perform the baseline UT examination at no later than 35 EFPY.

(9) Re-examination periods shall be determined by plant-specific evaluation per the MRP-227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [47] (i.e., 3% of baffle-former bolts with UT or visual indications or clustering* for downflow plants and 5% of baffle-former bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016-021 [48] and MRP 2017-009 [47]) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [51]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

• Clustering is defined per NSAL-16-1, Revision 1 [46] as three or more adjacent defective baffle-former bolts or more than 40% defective baffle-former bolts on the same baffle plate. Untestable bolts should be reviewed on a plant-specific basis consistent with WCAP-17096-NP-A for determination if these should be considered when evaluating clustering.

(10) See Westinghouse Technical Bulletin TB-19-5 dated 10/9/2019 and MRP 2019-017 dated 5/31/2019 for additional details on inspection recommendations.

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Westinghouse Non-Proprietary Class 3 C-6 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Primary Link Examination Expansion Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Control Rod Guide All plants Cracking (SCC, W2.CRGT Enhanced visual (EVT-1) A minimum of 75% of Tube Assembly Fatigue) Lower Flange examination to determine the the CRGT assembly W2.1.Remaining CRGT Aging Welds presence of crack-like surface lower flange weld lower flange welds Management (IE flaws in flange welds. surfaces and 0.25-inch and TE) Subsequent examination on a of the adjacent base 10-year interval. metal for the flange welds not inspected under the primary link.

See Figure A-3.

Bottom Mounted All plants Cracking (Fatigue) W2.CRGT Visual (VT-3) examination. 100% of BMI column Instrumentation including the Lower Flange Reinspection every 10 years bodies for which System detection of Welds following initial inspection. difficulty is detected W2.2.Bottom-mounted completely during flux thimble instrumentation (BMI) fractured column insertion/withdrawal.

column bodies bodies Aging See Figure A-9 and A-Management (IE) 11.

Core Barrel Assembly All plants Cracking (SCC) W3.Upper Core Enhanced visual (EVT-1) 100% of the accessible W3.1.Upper Girth Weld Barrel Flange examination. Reinspection weld length of one side (UGW) Weld (UFW) every 10 years following of the UGW and 3/4 of initial inspection. adjacent base metal shall be examined.

(Notes 2 and 5)

See Figure A-4.

WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 C-7 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Primary Link Examination Expansion Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Core Barrel Assembly All plants Cracking (SCC) W3.Upper Core Enhanced visual (EVT-1) 100% of the accessible W3.2.Upper Axial Weld Barrel Flange examination. Reinspection weld length of one side (UAW) Weld (UFW) every 10 years following of the UAW and 3/4 of initial inspection. adjacent base metal shall be examined.

(Notes 2 and 5)

See Figure A-4.

Core Barrel Assembly All plants Cracking (SCC) W3.Upper Core Enhanced visual (EVT-1) 100% of the accessible W3.3.Lower Flange Barrel Flange examination. Reinspection weld length of the OD Weld (LFW) Weld (UFW) every 10 years following surface of the LFW and initial inspection. 3/4 of adjacent base metal shall be examined.

(Note 5)

See Figure A-4.

Lower Internals All plants Cracking (SCC) W3.Upper Core Visual (VT-3) examination. Minimum of 25% of Assembly NOTE: Aging Barrel Flange Reinspection every 10 years bottom (non-core side)

W3.4.Lower support Management Weld (UFW) following initial inspection. surface (Note 3).

Catawba Unit forging or castings (TE in Casting) 1 and Unit 2 have a lower See Figure A-8 and A-9.

support forging WCAP-18496-NP September 2020 Revision 0

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Westinghouse Non-Proprietary Class 3 C-8 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Primary Link Examination Expansion Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Upper Internals All plants Cracking W4.Lower Girth Visual (VT-3) examination. Minimum of 25% of Assembly (Fatigue), Wear, Weld (LGW) Reinspection every 10 years core side surfaces W4.1.Upper core plate Aging following initial inspection. (Note 3).

Management (IE)

See Figure A-12.

Core Barrel Assembly All plants Cracking (SCC, W4.Lower Girth Enhanced visual (EVT-1) 100% of the accessible W4.2.Middle Axial IASCC) Weld (LGW) examination. Reinspection weld length of the OD of Welds (MAW) and Aging every 10 years following the MAW and LAW and W4.3.Lower Axial Management (IE) initial inspection 3/4of adjacent base Welds (LAW) metal shall be examined (Notes 5 and 6).

See Figure A-4.

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Westinghouse Non-Proprietary Class 3 C-9 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Primary Link Examination Expansion Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Lower Support All plants Cracking (IASCC) W4.Lower Girth Visual (VT-3) examination. 25% of the total number Assembly Aging Weld (LGW) Reinspection every 10 years of column assemblies W4.4.Lower support Management (IE) following initial inspection. (both visible and non-column bodies (both cast visible from above the and non-cast) lower core plate) using a VT-3 examination from above the lower core plate. The inspection coverage must be evenly distributed across the population of column assemblies.

(Notes 3 and 4)

See Figures A-8, A-9, and A-10.

Core Barrel Assembly All plants Cracking W6.Baffle- Volumetric (UT) 100% of accessible W6.1.Barrel-former (IASCC, Fatigue) former bolts (also examination. Reinspection barrel-former bolts bolts Aging refer to MRP every 10 years following (Minimum of 75% of the Management 2018-002) initial inspection. total population).

(IE, Void Swelling, Accessibility may be and ISR) limited by presence of thermal shield or neutron pads.

See Figure A-7.

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• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-10 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Primary Link Examination Expansion Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Lower Support All plants Cracking W6.Baffle- Volumetric (UT) 100% of accessible LSC Assembly (IASCC, Fatigue) former bolts examination. Reinspection bolts (Minimum of 75%

W6.2.Lower support Aging every 10 years following of the total population) column bolts Management initial inspection. or as supported by plant-(IE and ISR) specific justification.

See Figures A-8, A-9, and A-10.

Table C-2 Notes:

(1) Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table C-4.

(2) Examination coverage requires examination of either the ID or the OD of the weld.

(3) The stated minimum coverage requirement is the minimum if no significant indications are found. However, the Examination Acceptance criteria in Section 5 of MRP-227 1-A require that additional coverage must be achieved in the same outage if significant flaws are found. This contingency should be considered for inspection planning purposes.

(4) Justification that adequate distribution of the inspection coverage has been achieved can be based on geometric or layout arguments. Possible examples include, but are not limited to, inspection of all column assemblies in one quadrant of the lower core plate (based on the azimuthal symmetry of the plate) or inspecting every fourth column across the entire plate.

(5) A minimum coverage of 75% of the weld length on the surface being examined shall be achieved; however, for welds with limited access (Note 6), a minimum examination coverage of 50% of the weld length on the surface being examined shall be achieved.

(6) Accessibility to the MAW and LAW may be limited by the thermal shield or neutron panels - no disassembly to achieve higher weld length coverage is required.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-11 Table C-3: MRP-227, Revision 1-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage W10. Core Barrel All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Assembly (Wear) Section XI to determine general specified frequency.

Core barrel flange condition for excessive wear.

W11.Upper Internals All plants Cracking ASME Code Visual (VT-3) examination. All accessible surfaces at Assembly (SCC, Fatigue) Section XI specified frequency.

Upper support ring or skirt W12a.Lower Internals All plants Cracking ASME Code Visual (VT-3) examination All accessible surfaces at Assembly (IASCC, Fatigue) Section XI as of the lower core plates to specified frequency.

Lower core plate Aging supplemented detect evidence of distortion XL lower core plate Management (IE) by TB-16-4 and/or loss of bolt integrity.

(Note 1) (Note 3)

W12b.Lower Internals All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Assembly (Wear) Section XI as specified frequency.

Lower core plate supplemented XL lower core plate by TB-16-4 (Note 1) (Note 3)

W13.Bottom-Mounted All plants Loss of material IEB 88-09 Surface (ET) examination. Eddy current surface Instrumentation System (Wear) examination as defined in Flux thimble tubes plant response to IEB 88-09.

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• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-12 Table C-3: MRP-227, Revision 1-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage W14.Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing Components (TB-14-5) (Wear) Section XI as specified frequency.

Clevis bearing Stellite Cracking (SCC) supplemented wear surface by TB-14-5 (Note 2)

Clevis insert bolts (Note 2)

W15.Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing Components (Wear) Section XI as specified frequency.

Upper core plate supplemented alignment pins by TB-16-4 (Note 3)

Table C-3 Notes:

(1) XL = Extra Long, referring to Westinghouse plants with 14-foot cores. This component is not applicable to Catawba.

(2) The clevis inserts are attached to integrally welded reactor vessel lugs and the inserts are bolted to the lugs. The ASME Code examination of accessible surfaces is considered to include all details of the clevis configuration, including the bolting and locking devices. The bolting is fabricated from nickel-based materials and is susceptible to stress corrosion cracking (SCC). Although failure of the bolting does not itself cause loss of support function, asset impairment or issues with core barrel removal are a subsequent possibility. Westinghouse technical bulletin TB 14-5 dated 8/25/2014 provides additional information regarding possible visual indications that clevis bolting failure may have occurred. This information should be reviewed to ensure a heightened awareness of the examiners is applied to this Code inspection.

(3) Technical Bulletin TB-16-4 is not applicable to Catawba Unit 1 and Unit 2.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-13 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W1.Control Rod All plants Per the requirements of None N/A Per WCAP-17451-P.

Guide Tube WCAP-17451-P Assembly The specific relevant Guide plates condition is wear that (cards) could lead to loss of control rod alignment and impede control assembly insertion.

W2.Control Rod All plants Enhanced visual (EVT- W2.1.Remaining Confirmation of surface-breaking indications in For BMI column Guide Tube 1) examination. accessible CRGT two or more CRGT lower flange welds shall bodies, the specific Assembly The specific relevant lower flange welds require visual (EVT-1) examination of the relevant condition for Lower flange welds condition is a detectable remaining accessible CRGT lower flange welds the VT-3 examination crack-like surface and visual (VT-3) examination of BMI column is completely fractured W2.2.Bottom- bodies by the completion of the next refueling column bodies.

indication. mounted outage.

instrumentation (BMI) column bodies WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-14 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W3.Core Barrel All plants Periodic enhanced visual W3.1.Upper girth a. The confirmed detection and sizing of a The specific relevant Assembly (EVT-1) examination. weld (UGW) surface-breaking indication with a length condition for the Upper flange weld greater than 2 inches in the UFW shall require expansion core barrel (UFW) that the inspection be expanded to include the welds (UGW, LFW, The specific relevant W3.3.Lower flange UGW and LFW by the completion of the next UAW) and lower condition is a detectable weld (LFW) (Note 2) refueling outage. support forging or crack-like surface casting examinations is indication. b. The confirmed detection and sizing of a W3.2.Upper axial surface breaking indication with a length a detectable crack-like welds (UAW) greater than 2 inches in either the UGW or surface indication.

LFW shall require that the inspection be expanded to include the UAW by the W3.4.Lower support completion of the next refueling outage.

forging/casting

c. The confirmed detection of a surface-breaking indication with a length greater than 2 inches in the LFW shall require the inspection of the lower support forging or casting (25% of the non-core side surface) within the next 3 refueling outages. If an indication is found in this inspection of the lower support forging or casting, the examination coverage shall be expanded to 100% of the accessible surface of the noncore side surface of the lower support forging or casting during the same refueling outage.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-15 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W4.Core Barrel All plants Periodic enhanced visual W4.1.Upper core Plate a. The confirmed detection and sizing of a surface- a. The specific relevant Assembly (EVT-1) examination. breaking indication with a length greater than conditions for the Lower girth weld The specific relevant two inches in the LGW shall require inspection inspection of the W4.4.Lower support of the upper core plate (25% of the core-side upper core plate are (LGW) condition is a detectable column bodies (cast and crack-like surface surface) within the next 3 refueling outages. If broken or missing non-cast) an indication is found in this inspection of the parts of the plate.

indication.

upper core plate, the examination coverage shall b. The specific relevant W4.2.Middle axial be expanded to 100% of the accessible surface of conditions for the welds (MAW) the core-side surface of the upper core plate inspection of the during the same refueling outage. lower support

b. The confirmed detection and sizing of a surface- column bodies (cast W4.3.Lower axial breaking indication with a length greater than and non-cast) are welds (LAW) two inches in the LGW shall require inspection fractured, of the lower support column bodies (cast and misaligned, or non-cast) within the next 3 refueling outages. missing columns.

The confirmed detection of fractured, c. The specific relevant misaligned, or missing lower support columns condition for the shall require examination of 100% of the expansion MAW accessible uninspected lower support column and LAW assemblies using a VT-3 examination from inspections is a above the lower core plate (minimum of 75% of detectable crack-like the total population of lower support column surface indication.

assemblies) during the same outage.

c. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require that the inspections be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-16 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W5.Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly with baffle- examination.

Baffle-edge bolts edge bolts The specific relevant NOTE: conditions are missing or Not broken locking devices, applicable to failed or missing bolts, Catawba Unit and protrusion of bolt 1 and Unit 2 heads.

W6.Baffle-Former All plants Volumetric (UT) W6.2.Lower support Confirmation that more than 5% of the baffle The examination Assembly examination. column bolts former bolts actually examined on the four (4) acceptance criteria for Baffle-former bolts The examination baffle plates at the largest distance from the core the UT of the lower acceptance criteria for (presumed to be the lowest-dose locations) support column bolts W6.1.Barrel-former contain unacceptable indications shall require UT and the barrel-former the UT of the baffle- bolts (Note 3) former bolts shall be examination of the lower support column bolts bolts shall be established as part of the within the next 3 fuel cycles. established as part of examination technical the examination justification. technical justification.

Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require inspection of the barrel-former bolts within 3 refueling cycles.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-17 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W7.Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly examination.

Assembly The specific relevant (Includes: Baffle conditions are evidence plates, baffle edge of abnormal interaction bolts, corner bolts, with fuel assemblies, and indirect effects gaps along high fluence of void swelling in baffle plate joints, former plates) vertical displacement of baffle plates near high fluence joints, or more than 2 broken or damaged edge bolt locking systems along high fluence baffle plate joints.

W8.Alignment All plants Direct physical None N/A N/A and Interfacing with 304 measurement of spring Components stainless steel height.

Internals hold- hold-down The examination down spring springs acceptance criterion for NOTE: this measurement is that Not the remaining applicable to compressible height of Catawba Unit the spring shall provide 1 and Unit 2 hold-down forces within the plant-specific design tolerance.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-18 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Acceptance Criteria Examination Item Applicability (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria W9.Thermal All plants Visual (VT-3) None N/A N/A Shield Assembly with thermal examination.

Thermal shield shields The specific relevant flexures NOTE: conditions for thermal Not shield flexures are applicable to excessive wear, fracture, Catawba Unit or complete separation.

1 and Unit 2 Table C-4 Notes:

(1) The examination acceptance criterion for visual examination is the absence of the specified relevance condition(s).

(2) The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

(3) If significant baffle-former bolt clustering (as defined in MRP 2017-009 [47]) is discovered, Catawba will implement the Needed requirements of MRP 2018-002 [77] within 3 fuel cycles.

WCAP-18496-NP September 2020 Revision 0

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

WCAP-18496-NP Revision 0 Proprietary Class 3

• • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Zindren Taylor Sep-16-2020 14:44:05 Verifier Approval Mckinley Joshua K Sep-16-2020 15:22:07 Manager Approval Musser Kaitlyn M Sep-16-2020 16:37:41 Files approved on Sep-16-2020

• • • This record was final approved on 9/16/2020 4:37:41 PM. (This statement was added by the PRIME system upon its validation)

ATTACHMENT 2 Catawba Units 1 & 2 Summary Report for the Fuel Design/Management Assessments to Demonstrate MRP-227, Revision 1-A Applicability (Proprietary and Non-Proprietary)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 1 of 8 Attachment 2 Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Catawba Unit 1 and Unit 2 (Non-Proprietary)

© 2020 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 2 of 8 Introduction and Background This attachment summarizes the plant-specific assessment of the MRP-227, Revision 1-A (Reference 1) fuel design and fuel management limitations that was performed in support of the Catawba Unit 1 and Unit 2 Aging Management Plan (AMP) update.

Section 2.4 of MRP-227, Revision 1-A states, in part, that:

Users of these guidelines are expected to confirm with reasonable assurance that each reactor managed with the guidelines satisfies the assumptions discussed as follows. General assumptions used in the analysis include:

[no more than] 30 years of operation with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation, as well as the average core power levels and proximity of active fuel to the upper core support plate satisfies limits as described in Appendix B for Westinghouse/CE plants.

The following background information was taken from Appendix B of Reference 1:

The Safety Evaluation (SE) issued on Materials Reliability Program (MRP) technical report MRP-227, Revision 0, by the U.S. Nuclear Regulatory Commission (NRC) contained eight Applicant/Licensee Action Items (A/LAIs). These eight action items must be completed in the implementation of the Inspection and Evaluation (I&E) Guidelines outlined in MRP-227-A.

On November 28, 2012, a public meeting was held at the NRC office to discuss staff expectations and concerns regarding industry responses to A/LAIs 1 and 2. The concerns were addressed to owners of currently operating pressurized water reactor plants designed by Westinghouse and Combustion Engineering (CE). A series of proprietary and public meetings were conducted from January to June of 2013. At these meetings, the NRC, Westinghouse, the Electric Power Research Institute (EPRI), and utility representatives discussed regulatory concerns and determined a path for a comprehensive and consistent utility response to demonstrate applicability of MRP-227.

Westinghouse summarized the proprietary meeting presentations and supporting proprietary generic design basis information in WCAP-17780-P, and provided it to the NRC. WCAP-17780-P provides background proprietary design information regarding variances in stress, fluence, and temperature in the plants designed by Westinghouse and CE to support NRC reviews of utility submittals to demonstrate plant-specific applicability of MRP-227. NRC staff assessed this information provided in WCAP-17780-P and EPRI MRP-2013-025 as documented in ML14309A484.

Plant-specific evaluation to demonstrate the applicability of MRP-227 for managing aging would need to consider the following items:

1. designated design specific criteria in responding to specific NRC requests for additional information,

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 3 of 8

2. criteria defined in MRP-227, Section 2.4, and

3. plant-specific regulatory commitments for managing aging in reactor internals.

The NRC staff subsequently stated that the information provided by the industry to the NRC staff demonstrated that the MRP-227 I&E Guidelines are applicable for the range of conditions expected at the currently operating Westinghouse and CE-designed plants in the U.S.

As a result of the technical discussions with the NRC staff, the basis for a plant to respond to the NRCs Request for Additional Information (RAI) to demonstrate compliance with MRP-227 for originally licensed and uprated conditions was determined to be satisfied with plant-specific responses to the following question related to plant-specific Fuel Design and/or Fuel Management:

Question: Does the plant have atypical fuel design or fuel management that could render the assumptions of MRP-227, regarding core loading/core design, non-representative for that plant?

Appendix B of Reference 1 provides guidance for demonstrating compliance with the fuel design and fuel management limitations associated with this question. The plant-specific assessment of these limitations for Catawba Units 1 and 2 was performed in accordance with this guidance.

Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations Catawba Units 1 and 2 have not utilized atypical fuel design or fuel management that could make the assumptions of MRP-227, Revision 1-A regarding core loading/core design non-representative, including power changes/uprates that have occurred over their operating lifetimes. This conclusion is based on a comparison of the Catawba Unit 1 and 2 core geometries and operating characteristics with the applicability guidelines for Westinghouse-designed reactors specified in Appendix B of MRP-227, Revision 1-A.

2.1 Radial Boundary Limitations Appendix B of Reference 1 states that:

The primary driver for the radial core power distribution is the MRP-227 basis of 30 years of out-in management, where fresh fuel is placed in peripheral core locations, followed by 30 years of low leakage fuel management. Any change in this scenario has the potential to impact re-inspection, but not the initial inspection timing or affected components. Due to design similarities across the currently operating Westinghouse and CE U.S. fleet, in most, but not all cases, internals component geometry is a secondary effect. Neutron flux and heating rate could be expected to vary by as much as a factor of 5, depending on radial core power distribution and absolute rated power.

Local effects at key locations are dominated by a few (typically 3-5) fuel assemblies located on the core periphery. There is no impact on the initial inspection, but a change from the low leakage operating characteristics during the second 30 years of operation could impact the MRP-227 re-

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 4 of 8 inspection recommendations. To provide assurance that there would not be higher than anticipated rates of degradation in the later years of operation, MRP-227 guidance on applicability of the recommendations precludes return to out-in core loading patterns. The limitations on power for demonstrating applicability in the peripheral assemblies provided in this guideline preclude return to the more damaging out-in core loading pattern.

Comparison: Catawba Unit 1 implemented a low-leakage fuel management strategy with the 10th fuel cycle following 11.0 calendar years (9.2 effective full-power years (EFPY)) of operation and has been implementing low-leakage core designs since that time. There are currently no plans to return to out-in fuel management.

Catawba Unit 2 implemented a low-leakage fuel management strategy with the 11th fuel cycle following 13.6 calendar years (11.9 EFPY) of operation and has been implementing low-leakage core designs since that time. There are currently no plans to return to out-in fuel management.

In addition to precluding a return to out-in core loading patterns, Appendix B of Reference 1 states that plant-specific applicability of MRP-227 in the radial direction with no further evaluation required is demonstrated by meeting the following limits:

Limit 1: For operation going forward, the nuclear heat generation rate figure of merit (HGR-FOM) shall be less than or equal to 68 W/cm3.

Comparison: For the last five operating fuel cycles (Cycles 20-24) at Catawba Unit 1, the HGR-FOM at key baffle locations has ranged between [ ]a,c. This range of HGR-FOMs is representative of anticipated future operation.

For the last five operating fuel cycles (Cycles 19-23) at Catawba Unit 2, the HGR-FOM at key baffle locations has ranged between [ ]a,c. This range of HGR-FOMs is representative of anticipated future operation.

Limit 2: For operation going forward, the average power density of the reactor core shall be less than 124 W/cm3.

Comparison: For the last five operating fuel cycles (Cycles 20-24), Catawba Unit 1 has been operating at the following power levels (core power densities for 193 fuel assembly geometry):

• Cycles 20-22: 3411 MWt (104.49 W/cm3)

• Cycles 23-24: 3469 MWt (106.27 W/cm3)

The power level of 3469 MWt is representative of anticipated future operation of Catawba Unit 1.

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 5 of 8 For the last five operating fuel cycles (Cycles 19-23), Catawba Unit 2 has been operating at the following power level (core power density for 193 fuel assembly geometry):

• Cycles 19-23: 3411 MWt (104.49 W/cm3)

The power level of 3411 MWt is representative of anticipated future operation of Catawba Unit 2.

2.2 Upper Axial Boundary Limitations Appendix B of Reference 1 states that plant-specific applicability of MRP-227 in the upper axial direction with no further evaluation required is demonstrated by meeting the following limits:

Limit 1: Considering the entire operating lifetime of the reactor, the distance between the top of the active fuel stack and the bottom of the upper core plate (UCP) shall not be less than or equal to 12.2 inches for a period of more than two years.

Comparison: For the Catawba Unit 1 reactor internals and fuel assembly geometry, the nominal distance between the top of the active fuel stack and the bottom of the upper core plate (UCP) averaged over the first 24 fuel cycles of operation was [ ]a,c. During that period of time, the nominal distance between the top of the active fuel and the bottom of the UCP and was less than 12.2 inches for an operating period of more than 7 years. More specifically, during Cycles 7-12, the nominal distance between the top of the active fuel stack and the bottom of the UCP ranged from [ ]a,c.

A fuel-assembly-to-UCP gap of less than 12.2 inches for an operating period greater than two years violates the established MRP-227, Revision 1-A criterion. However, a more detailed evaluation performed for Catawba Unit 1 demonstrated that the increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-12 was more than offset by the margin afforded by the lower operating power density over the entire plant lifetime.

In particular, this evaluation demonstrated that the core power densities for all Catawba Unit 1 fuel cycles were less than the 124.0 W/cm3 criterion by approximately [

]a,c. Since neutron exposure rates are directly proportional to reactor power, these lower, plant-specific core power densities result in a [ ]a,c reduction in neutron exposure rates relative to those that would occur during operation at 124.0 W/cm3.

Conversely, the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-12 were estimated to result in a [ ]a,c increase in neutron exposure rates relative to those that would occur during operation at the 12.2-inch gap criterion. For all other fuel cycles, the fuel-assembly-to-UCP gap criterion was met with associated margins in the resulting neutron exposure rates ranging from [ ]a,c.

When the entire operating period of Catawba Unit 1 is considered, the significantly larger margins associated with the plant-specific core power densities more than offset the

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 6 of 8 increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-12. Therefore, it can be concluded that the neutron exposure rates for the Catawba Unit 1 components located above the reactor core are less than those that would occur during operation with a core power density and fuel-assembly-to-UCP distance equal to the MRP-227, Revision 1-A criteria.

For the Catawba Unit 2 reactor internals and fuel assembly geometry, the nominal distance between the top of the active fuel stack and the bottom of the upper core plate (UCP) averaged over the first 23 fuel cycles of operation was [ ]a,c. During that period of time, the nominal distance between the top of the active fuel and the bottom of the UCP and was less than 12.2 inches for an operating period of more than 5 years. More specifically, during Cycles 7-10, the nominal distance between the top of the active fuel stack and the bottom of the UCP ranged from [ ]a,c.

A fuel-assembly-to-UCP gap of less than 12.2 inches for an operating period greater than two years violates the established MRP-227, Revision 1-A criterion. However, a more detailed evaluation performed for Catawba Unit 2 demonstrated that the increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-10 was more than offset by the margin afforded by the lower operating power density over the entire plant lifetime.

In particular, this evaluation demonstrated that the core power densities for all Catawba Unit 2 fuel cycles were less than the 124.0 W/cm3 criterion by approximately [ ]a,c.

Since neutron exposure rates are directly proportional to reactor power, these lower, plant-specific core power densities result in a [ ]a,c reduction in neutron exposure rates relative to those that would occur during operation at 124.0 W/cm3. Conversely, the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-10 were estimated to result in a

[ ]a,c increase in neutron exposure rates relative to those that would occur during operation at the 12.2-inch gap criterion. For all other fuel cycles, the fuel-assembly-to-UCP gap criterion was met with associated margins in the resulting neutron exposure rates ranging from [ ]a,c.

When the entire operating period of Catawba Unit 2 is considered, the significantly larger margins associated with the plant-specific core power density more than offset the increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 7-10. Therefore, it can be concluded that the neutron exposure rates for the Catawba Unit 2 components located above the reactor core are less than those that would occur during operation with a core power density and fuel-assembly-to-UCP distance equal to the MRP-227, Revision 1-A criteria.

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 7 of 8 Limit 2: Considering the entire operating lifetime of the reactor, the average power density of the core shall not be greater than or equal to 124 W/cm3 for a period of more than two years.

Comparison: Over the operating lifetime of Catawba Unit 1, the rated core power level, including power uprates, has increased from 3411 MWt to 3469 MWt. This variation of rated power level corresponds to a power density range of 104.49 W/cm3 to 106.27 W/cm3.

Over the operating lifetime of Catawba Unit 2 reactor, the rated core power level has been 3411 MWt. This rated power level corresponds to a power density of 104.49 W/cm3.

2.3 Lower Axial Boundary Limitations Appendix B of Reference 1 states that plant-specific applicability of MRP-227 in the lower axial direction with no further evaluation required is demonstrated by meeting the criteria in Section 2.4 of MRP-227, Revision 1-A.

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Attachment 2 of LTR-REA-19-131, Revision 0 January 28, 2020 Page 8 of 8 References

1. EPRI Technical Report, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), EPRI, Palo Alto, CA: 2019. 3002005349.

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

LTR-REA-19-131 Revision 0 Proprietary Class 2 (R)

• • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Hawk Andrew E Jan-28-2020 14:57:22 Reviewer Approval Hayes Eugene Jan-28-2020 15:13:48 Manager Approval Houssay Laurent Jan-28-2020 15:24:54 Files approved on Jan-28-2020

• • • This record was final approved on 1/28/2020 3:24:54 PM. (This statement was added by the PRIME system upon its validation)

ATTACHMENT 3 Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-20-5005, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice

Westinghouse Non-Proprietary Class 3 CAW-20-5005 Page 1 of3 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Camille T. Zozula, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions ofLTR-REA-19-131 Attachment 1, Revision Obe withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses.

Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 CAW-20-5005 Page 2 of 3 AFFIDAVIT (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(t) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means oflower case letters (a) through (t) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These

Westinghouse Non-Proprietary Class 3 CAW-20-5005 Page 3 of 3 AFFIDAVIT lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (t) of this Affidavit.

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: Olo\<)JWW Infrastructure & Facilities Licensing