RA-23-0218, Review Request for the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals

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Review Request for the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals
ML23264A034
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 09/21/2023
From: Haaf T
Duke Energy Progress
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-23-0218 WCAP-18710-NP
Download: ML23264A034 (1)
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Text

Thomas P. Haaf Site Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill, NC 27562-9300 984-229-2512 September 21, 2023 Serial: RA-23-0218 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

Review Request for the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals Ladies and Gentlemen:

In accordance with the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Updated Final Safety Analysis Report (UFSAR), Section 18.1, Duke Energy Progress, LLC (Duke Energy), is providing the HNP Reactor Vessel Internals (RVI) Aging Management Program (AMP) and Inspection Plan. The HNP UFSAR, Section 18.1 states that, In accordance with the guidance of NUREG-1801 [Generic Aging Lessons Learned (GALL) Report], regarding aging management of reactor vessel internals components for aging mechanisms, such as embrittlement and void swelling, HNP will: (1) participate in the industry programs for investigating and managing aging effects on reactor internals (such as Westinghouse Owner's Group and Electric Power Research Institute materials programs), (2) evaluate and implement the results of the industry programs as applicable to the reactor internals, and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval. contains the AMP and Inspection Plan for the HNP RVI. The AMP and Inspection Plan is based upon the Electric Power Research Institute Technical Report MRP-227, Rev. 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19339G350), which provides a strategy for managing age-related material degradation in RVI components through the period of extended operation. Enclosure 2 contains the plant-specific assessment of the MRP-227, Revision 1-A, fuel design and fuel management limitations for HNP.

Section 6 of Enclosure 1 describes how HNP meets the applicability requirements contained in MRP-227, Revision 1-A. Appendix B of MRP-227, Revision 1-A, provides guidance on plant-specific fuel design and fuel management requirements. Enclosure 2 addresses Appendix B of MRP-227, Revision 1-A, and supplements the information contained in Section 6 of Enclosure

1.

Duke Energy is making a commitment to update the HNP UFSAR with a description of the AMP and Inspection Plan for the HNP RVI, which is identified in Enclosure 3.

U.S. Nuclear Regulatory Commission RA-23-0218 Page2 Duke Energy requests approval of this submittal by September 30, 2024, to support MRP-227 implementation activities. Please refer any questions regarding this submittal to Sarah McDaniel, HNP Regulatory Affairs, at (984) 229-2002.

Enclosures:

1. WCAP-18710-NP, Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals, Revision 0
2. CQL-REAC-TM-AA-000001, Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris, Revision 0
3. Regulatory Commitment cc:

P. Boguszewski, NRC Senior Resident Inspector, HNP M. Mahoney, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission RA-23-0218 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 WCAP-18710-NP, Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals, Revision 0

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-18710-NP August 2023 Revision 0 Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 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

© 2023 Westinghouse Electric Company LLC All Rights Reserved WCAP-18710-NP_Revision_0.docx-080723 WCAP-18710-NP Revision 0 Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals Micah D. Young*

Materials and Aging Management August 2023 Verifier:

Louis W. Turicik*

Materials and Aging Management Reviewer:

Joshua K. McKinley*

Materials and Aging Management Approved: Kaitlyn M. Musser*, Manager Materials and Aging Management

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii WCAP-18710-NP August 2023 Revision 0 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 HARRIS REACTOR VESSEL INTERNALS................................................................. 2-2 2.3 HARRIS LICENSE RENEWAL..................................................................................... 2-4 3

PROGRAM OWNER................................................................................................................... 3-1 4

DESCRIPTION OF HARRIS REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS........................................................................... 4-1 4.1 HARRIS PROGRAMS AND ACTIVITIES SUPPORTING AGING MANAGEMENT OF RVI............................................................................................... 4-4 4.1.1 Water Chemistry Program............................................................................... 4-4 4.1.2 ASME Section XI Program............................................................................. 4-4 4.1.3 Flux Thimble Tube Inspection Program.......................................................... 4-5 4.1.4 Upflow Design Reactor Vessel Internals......................................................... 4-5 4.1.5 Control Rod Guide Tube Split Pin Replacement Project................................. 4-6 4.1.6 Reactor Vessel Head and Thermal Sleeve Replacement.................................. 4-6 4.1.7 Power Uprate Project....................................................................................... 4-6 4.2 INDUSTRY PROGRAMS............................................................................................... 4-7 4.2.1 Aging Management for Reactor Internals (WCAP-14577)............................. 4-7 4.2.2 Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A).................................................................................................. 4-7 4.2.3 Reactor Internals Acceptance Criteria Methodology and Data Requirements (WCAP-17096-NP-A).............................................................. 4-9 4.2.4 Reactor Internals Guide Tube Wear (WCAP-17451-P)................................... 4-9 4.2.5 Core Barrel Operating Experience................................................................ 4-10 4.2.6 Baffle-Former Bolt Degradation (NSAL-16-1)............................................. 4-11 4.2.7 Clevis Bearing STELLITE Wear Surface and Clevis Insert Bolts (TB-14-5).............................................................................................................. 4-12 4.2.8 Thermal Sleeve Flange Wear......................................................................... 4-12 4.2.9 On-Going Industry Programs and NEI 03-08 Guidelines............................. 4-13 4.3

SUMMARY

................................................................................................................... 4-14 5

SHEARON HARRIS REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES.............................................................................................................................. 5-1 5.1 GALL ELEMENT 1: SCOPE OF PROGRAM............................................................... 5-2 5.2 GALL ELEMENT 2: PREVENTATIVE ACTIONS....................................................... 5-4 5.3 GALL ELEMENT 3: PARAMETERS MONITORED OR INSPECTED...................... 5-5 5.4 GALL ELEMENT 4: DETECTION OF AGING EFFECTS........................................... 5-7 5.5 GALL ELEMENT 5: MONITORING AND TRENDING............................................ 5-10

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii WCAP-18710-NP August 2023 Revision 0 5.6 GALL ELEMENT 6: ACCEPTANCE CRITERIA....................................................... 5-12 5.7 GALL ELEMENT 7: CORRECTIVE ACTIONS......................................................... 5-13 5.8 GALL ELEMENT 8: CONFIRMATION PROCESS.................................................... 5-14 5.9 GALL ELEMENT 9: ADMINISTRATIVE CONTROLS............................................. 5-15 5.10 GALL ELEMENT 10: OPERATING EXPERIENCE................................................... 5-16 6

MRP-227 SAFETY EVALUATION CONDITIONS AND ACTION ITEMS............................. 6-1 6.1 MRP-227, REVISION 1-A GUIDELINE APPLICABILITY......................................... 6-1 6.2 MRP-227, REVISION 1 SE APPLICANT/LICENSEE ACTION ITEM 1:

DEGRADATION OF BAFFLE-FORMER BOLTS........................................................ 6-6 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 SHEARON HARRIS UNIT 1 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

................................................................................................................. B-1 APPENDIX C MRP-227 AUGMENTED INSPECTIONS.................................................................... C-1

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv WCAP-18710-NP August 2023 Revision 0 LIST OF TABLES Table 7-1 Shearon Harris Unit 1 Primary Component Inspection Plan..................................................... 7-2 Table B-1: Summary of Aging Management Evaluation - Internals........................................................ 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-7 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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v

WCAP-18710-NP August 2023 Revision 0 LIST OF FIGURES Figure A-1: Illustration of Harris Internals.............................................................................................. A-1 Figure A-2: Harris Control Rod Guide Cards.......................................................................................... A-2 Figure A-3: Harris Control Rod Guide Tube Assembly........................................................................... A-3 Figure A-4: Harris Core Barrel Welds...................................................................................................... A-4 Figure A-5: Harris Baffle Plates (Octant-Symmetric).............................................................................. A-5 Figure A-6: Harris Core Baffle/Barrel Structure...................................................................................... A-6 Figure A-7: Harris Baffle-Former Structure............................................................................................ A-7 Figure A-8: Harris Lower Core Support Structure................................................................................... A-8 Figure A-9: Harris Lower Core Support Structure - Core Support Forging Cross-Section.................... A-9 Figure A-10: Harris Core Support Column............................................................................................ A-10 Figure A-11: Harris Bottom-Mounted Instrumentation (BMI) Column Design.................................... A-11 Figure A-12: Harris Upper Internals Assembly..................................................................................... A-12

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi WCAP-18710-NP August 2023 Revision 0 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 (ASME) Boiler & Pressure Vessel (Code)

B&W Babcock & Wilcox BFB baffle-former bolt 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 EC engineering change EFPY effective full-power years ET eddy current testing or examination EPRI Electric Power Research Institute EVT enhanced visual testing or examination (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 IC irradiation creep ID inner diameter IE irradiation embrittlement INPO Institute of Nuclear Power Operations ISI in-service inspection ISR irradiation-enhanced stress relaxation LAR License Amendment Request LAW lower axial weld LCP lower core plate LFW lower flange weld LGW lower girth weld LOCA loss-of-coolant accident LRA License Renewal Application LSC lower support column MAW middle axial weld MRP (EPRI) Materials Reliability Program MWt megawatts thermal NDE non-destructive examination NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vii WCAP-18710-NP August 2023 Revision 0 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 (formerly WOG)

PWSCC primary water stress corrosion cracking QA quality assurance RCCA reactor control cluster assembly RCS reactor coolant system RV reactor vessel RVCH reactor vessel closure head RVI reactor vessel internals SCC stress corrosion cracking SCs structures and components SE Safety Evaluation SER Safety Evaluation Report SG steam generator SLR Subsequent License Renewal SS stainless steel TE thermal embrittlement TLAA time-limited aging analysis TR topical report UAW upper axial weld UCP upper core plate UFW upper flange weld UFSAR Updated Final Safety Analysis Report UGW upper girth weld USC upper support column UT ultrasonic testing or examination (a volumetric NDE method)

VS void swelling VT visual testing or examination (a visual NDE method that includes VT-1 and VT-3)

WCAP Westinghouse Commercial Atomic Power (Report)

WOG Westinghouse Owners Group XL Extra Long

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 viii WCAP-18710-NP August 2023 Revision 0 ACKNOWLEDGEMENTS Gratitude is given to Rachel Doss, who manages the Fleet Reactor Vessel Internals at Duke Energy, for her support in the development of this report.

This report is the result of the combined efforts of a multidisciplinary team at Westinghouse. The completion of this report was realized through the contributions of the following individuals:

David Kovacic Sean Leskovic Jeffery Liberatore Joshua McKinley Kaitlyn Musser Louis Turicik Micah Young Taylor Zindren EP-RCCA is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates, and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

Inconel, Stellite, and Harmoni are trademarks or registered trademarks of their respective owners. Other names and trademarks may be trademarks of their respective owners

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 WCAP-18710-NP August 2023 Revision 0 1

PURPOSE The purpose of this report is to document the Duke Energy Shearon Harris Unit 1 (hereafter referred to as Harris) Reactor Vessel Internals (RVI) Aging Management Program (AMP) and Inspection Plan for submittal to the U.S. Nuclear Regulatory Commission (NRC). The purpose of the Harris RVI Program is to manage the effects of aging on RVI through the period of extended operations, which begins at midnight on October 25, 2026 [1]. This document demonstrates that the Harris RVI Program manages the effects of aging on RVI components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function throughout the period of extended operation at Harris. This document is supported by existing Duke Energy documents and procedures. As required by industry experience or directive in the future, the Harris 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 Harris will continue to be conducted in accordance with the current licensing basis (CLB) for the RVI by fulfilling license renewal

[2], American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI Inservice Inspection (ISI) programs [3, 4, 5, 32] (and the applicable Section XI Edition and Addenda

[6]), and industry requirements [7]. The Harris 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 Groups (PWROG).

The main objectives for the Harris 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 [8].

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

Define the Harris RVI Program, based on industry-defined (EPRI/MRP and PWROG) RVI requirements and guidance, addressing the ten AMP elements in NUREG-1801, Revision 2 [9] as updated by SLR-ISG-2021-01 [10].

Address the applicable Applicant/Licensee Action Item (A/LAI) identified in the Safety Evaluation (SE) on MRP-227, Revision 1-A (contained within MRP-227, Revision 1-A) [11], 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 Harris reactor internals.

Appendix A of the NRC SE [2] on the Harris License Renewal Application (LRA) [1] contains commitments for license renewal. In commitment 1, Duke Energy stated that they would participate in industry programs for investigating and managing aging effects on reactor internals, citing the Westinghouse Owners Group (formerly WOG, now PWROG) and EPRI programs (such as MRP), as examples. Duke Energy committed to evaluate and implement the results of the industry programs as applicable to the RVI; and, upon completion of these programs, submit an inspection plan for the RVI to the NRC for review and approval no later than 24 months before entering the extended period of operation.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-2 WCAP-18710-NP August 2023 Revision 0 In keeping with this commitment, Duke Energy developed a plant-specific RVI inspection plan for Harris to implement MRP-227, Revision 1-A.

This document supersedes Westinghouse Commercial Atomic Power (WCAP) report WCAP-17101-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Harris Nuclear Plant, [12].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 WCAP-18710-NP August 2023 Revision 0 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 - Final Report [13]. 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, Harris has defined its RVI and has demonstrated that the aging effects associated with the Harris 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 the 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 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 [14]. 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 internal 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 [15, 16, 17]:

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 postulated aging mechanisms (further discussed in Section 4 of this document).

Categorization of PWR internals components into groups, based on the screening criteria and the likelihood and severity of safety and economic consequences: components with insignificant effects from aging degradation, components with the potential to have moderately significant effects from aging degradations, 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 include component accessibility, operating experience (OE), existing evaluations, and prior examination results [18].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-2 WCAP-18710-NP August 2023 Revision 0 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 [11] (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, CE, and B&W). The document was submitted to the NRC for a formal evaluation and review. The Safety Evaluation Report (SER) was issued on April 25, 2019 [19].

MRP-228, Revision 4 [20], Inspection Standard for Pressurized Water Reactor Internals - 2020 Update, provides guidance on the qualification/demonstration of the required NDE techniques and other criteria pertaining to the actual performance of the inspections. This is the most recent revision of MRP-228 is at the time of this report; the revision of MRP-228 in effect at the time of the inspection should be utilized for future inspections.

Additionally, the industry has developed many other documents to provide guidance and tools for the aging management of reactor internals alongside and in support of MRP-227, Revision 1-A.

2.2 HARRIS REACTOR VESSEL INTERNALS The RVI for Harris are integral with the reactor coolant system (RCS) of a Westinghouse three-loop nuclear steam supply system (NSSS). Illustrations of typical RVI are provided in Figure A-1 through Figure A-12.

As described in the LRA [1], the RVI components are divided into three parts consisting of the lower core support structure, the upper core support structure, and the incore instrumentation support structure. The RVI support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and control rod drive mechanisms, direct reactor coolant flow past the fuel elements, direct reactor coolant flow to the pressure vessel head, provide gamma and neutron shielding, and guides for the incore instrumentation.

The major support member of the RVI is the lower core support structure. This support structure assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the neutron shield pads, and the core support (which is welded to the core barrel). All the major material for this structure is Type 304 stainless steel (SS). The lower core support structure is supported at its upper flange from a ledge in the reactor vessel (RV) flange and its lower end is restrained in its transverse movement by a radial support system attached to the vessel wall. Within the core barrel are a baffle assembly and a lower core plate, both of which are attached to the core barrel wall and enclose the periphery of the assembled core.

The support columns constitute the primary structural support for the lower core plate. The lower core support structure and core barrel serve to provide passageways and control for the reactor 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 for each assembly) are also inserted into this plate. The support columns between the lower core plate and

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-3 WCAP-18710-NP August 2023 Revision 0 the 16 in. thick core support casting, which forms the bottom of the core barrel, provide stiffness to the assembly and transmit the core load to the core support.

The neutron shield pad assembly consists of four pads that are bolted and pinned to the outside of the core barrel. These pads are constructed of Type 304 SS. Specimen guides in which RV material surveillance samples can be inserted and irradiated during reactor operation are attached to the pads.

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. The main radial support system of the lower end of the core barrel is accomplished by key and keyway joints to the RV wall. At equally spaced points around the circumference, an INCONEL clevis block is welded to the vessel inner diameter. Another INCONEL insert block is bolted to each of these blocks and has a keyway geometry. Opposite each of these is a key which is attached to the internals.

At assembly, as the internals are lowered into the vessel, the keys engage the keyways in the axial direction.

In the event of an abnormal downward vertical displacement of the internals following a hypothetical complete failure of the core barrel, energy absorbing devices installed at the bottom of the secondary core support structure limit the displacement after contacting the vessel bottom head. The secondary core support structure is located at the bottom of the core barrel assembly below the lower core support. The load from the hypothetical core drop is transferred through the energy absorbing devices of the internals to the vessel. The energy absorbers, cylindrical in shape, are contoured on their bottom surface to the RV bottom head geometry. Assuming a downward vertical displacement, the potential energy of the system is absorbed mostly by the strain energy of the energy absorbing devices.

The upper core support assembly consists of the upper support plate assembly and the upper core plate, between which are contained support columns and control rod guide tube assemblies. The 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 support columns transmit the mechanical loadings between the two plates and serve the supplementary function of supporting thermocouple guide tubes. The control rod guide tube assemblies sheathe and guide the control rod drive shafts and control rods. They 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 assembly 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 slots in the upper core plate.

As the upper core support assembly is lowered into the lower internals, the slots in the plate engage the flat-sided pins in the axial direction. Lateral displacement of the plate and of the upper core support assembly is restricted by this design. Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper core support assembly is lowered into place. Proper alignment of the lower core support structure, the upper core support assembly, the fuel assemblies, and control rods is thereby assured by this system of locating pins and guidance arrangement. The upper core support assembly is restrained from any axial movements by a large circumferential spring which rests between the core barrel flange and the upper core support assembly and is compressed by the RV head flange.

The incore instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom. The upper and lower systems utilize the RV head penetrations. The thermocouple conduits are supported from the columns of the upper core support system.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-4 WCAP-18710-NP August 2023 Revision 0 The thermocouple conduits are sealed stainless steel tubes. In addition to the upper incore instrumentation, there are bottom-mounted instrumentation (BMI) columns which carry the retractable, cold worked stainless steel flux thimbles that are pushed upward into the reactor core.

2.3 HARRIS LICENSE RENEWAL Duke Energy submitted an LRA for a renewed operating license for Harris [1]. In the SER (NUREG-1916

[2]) 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 Harris RVI components and subcomponents that are subject to AMP requirements is in Table 3.1.2-1 of the LRA [1]. In the SER, the NRC concluded that the Duke Energy 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 Harris reactor vessel internals components and subcomponents and flux thimble tubes already reviewed by the NRC is included in Table B-1. Included in the SER Appendix A is Commitment 1 associated with the RVI Inspection, and Commitment 4 associated with the management of the thermal aging and neutron irradiation embrittlement of cast austenitic stainless steel (CASS) program implementation. The existing commitments are also contained in Section 18.1 of the Harris Updated Final Safety Analysis Report (UFSAR) [23].

As part of its license renewal, Duke Energy stated that they would participate in industry activities associated with RVI-related issues and that the Harris 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 Harris RVI Program. The Harris RVI AMP and Inspection Plan, as documented herein, is based on MRP-227, Revision 1-A.

Commitment 4 of the SER states, The Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program is a new program to be implemented. This program was applied to the upper support column (USC) spider (USC bases) and bottom-mounted instrumentation (BMI) column cruciforms. These internals components should no longer be managed by the Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steels Program since the requirements have been subsumed into the RVI AMP herein. This comports with SLR-ISG-2021-01 [10], which uses the PWR Vessel Internals AMP to manage CASS internals components. Therefore, these components will be managed under the Harris RVI AMP and this commitment will be fulfilled.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-1 WCAP-18710-NP August 2023 Revision 0 3

PROGRAM OWNER The successful implementation and comprehensive long-term management of the Harris RVI program will require the integration of Duke Energy organizations, corporately and at Harris, 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 Nuclear Energy Institute (NEI) document NEI 03-08 [24] practices.

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

Roles and responsibilities for establishing, maintaining, and implementing the Harris RVI Program are established in the applicable Duke Energy administrative and program procedures [25].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-1 WCAP-18710-NP August 2023 Revision 0 4

DESCRIPTION OF HARRIS 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 Harris RVI program is to ensure the long-term integrity and safe operation of the RVI components. Harris developed this AMP and Inspection Plan in conformance with NUREG-1801, Revision 2 [9] (as updated by SLR-ISG-2021-01 [10]) and MRP-227, Revision 1-A

[11].

This RVI program utilizes a combination of prevention, mitigation, and condition monitoring. Where applicable, credit is taken for existing programs, such as water chemistry [26, 27], inspections prescribed by the ASME Section XI ISI Program [3, 4, 5, 32] and flux thimble tube inspection program [29], and past and future mitigation projects such as the control rod guide tube split pin replacement project [39] and reactor vessel head and thermal sleeve replacement [43]. 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 [1]. The 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 degradation mechanisms of concern. Therefore, this AMP and Inspection Plan is consistent with the existing Harris 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 material aging degradation mechanisms identified as affecting the 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.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2 WCAP-18710-NP August 2023 Revision 0 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 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.

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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-3 WCAP-18710-NP August 2023 Revision 0 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 Harris RVI incorporates programs and activities that are credited for managing the aging effects produced by the aging degradation mechanisms listed above.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-4 WCAP-18710-NP August 2023 Revision 0 4.1 HARRIS PROGRAMS AND ACTIVITIES SUPPORTING AGING MANAGEMENT OF RVI Harris has a number of programs and activities supporting the aging management of the RVI; these include:

Water Chemistry Program ASME Section XI ISI Program Flux Thimble Tube Inspection Program Upflow Design Reactor Vessel Internals Control Rod Guide Tube Split Pin Replacement Project Reactor Vessel Head and Thermal Sleeve Replacement Power Uprate Project Brief descriptions of the programs and activities are included in the following subsections.

4.1.1 Water Chemistry Program The Harris Water Chemistry Program [26, 27] is an existing program that provides activities for monitoring and controlling the chemical environments of the Harris 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 Harris.

The Harris Water Chemistry Program includes periodic sampling of primary water for the known detrimental contaminants specified in the EPRI PWR water chemistry guidelines [31] 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 Harris-specific procedures [26, 27].

Harris follows the guidance set forth in the EPRI PWR Primary Water Chemistry Guidelines [31]. The limits imposed by the Harris Program meet the intent of the industry standard for addressing primary water chemistry [26, 27].

4.1.2 ASME Section XI Program The Harris ASME Section XI Program [3, 4, 5, 32] 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 [32].

Table 4-9 considers the ASME Code requirements for VT-3 sufficient to monitor the applicable aging effects for the Existing Programs components, with the exception of the clevis bearing STELLITE' wear surface and clevis insert bolts. The ASME Code required exam for the clevis bearing STELLITE wear surface and clevis insert bolts is supplemented by TB-14-5 [33]. Harris will follow the ASME Code

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-5 WCAP-18710-NP August 2023 Revision 0 requirements for VT-3 examination and will follow TB-14-5 from the clevis bearing STELLITE wear surface and clevis insert bolts. Note that TB-16-4 [34] is not applicable to Harris.

4.1.3 Flux Thimble Tube Inspection Program 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 BMI column assemblies 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 Harris Flux Thimble Tube Inspection Program is a program that manages loss of material due to wear of the flux thimble tube materials [29]. It implements the recommendations of NRC Bulletin 88-09 [35]

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 Flux 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 [36]. 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 ET examination as defined in the plants response to NRC Bulletin 88-09, is considered sufficient to monitor for the applicable aging effect.

Note that the flux thimble tubes are a pressure boundary, rather than an RVI component. The flux thimble path from the seal table to the bottom-mounted nozzles is defined by the flux thimble guide tubes, which are part of the primary pressure boundary and not considered to be part of the RVI. Though they are not part of the RVI, the flux thimble tubes are included in this report due to their inclusion in MRP-227, Revision 1-A.

4.1.4 Upflow Design Reactor Vessel Internals Harris was originally constructed as a downflow unit. Plants with a downflow configuration are susceptible to baffle jetting and require a greater number of intact baffle-former bolts to assure the safety of the plant than does a comparable upflow plant.

In 1983 (prior to first startup), Harris swapped out the upper and lower internals with the canceled Harris Unit 3 internals. The design change switched Unit 1 internals from a downflow design to a standard upflow design [37]. This change reduced the plants susceptibility to baffle bolt cracking and the consequence of any such bolt failure during a postulated loss-of-coolant accident (LOCA).

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-6 WCAP-18710-NP August 2023 Revision 0 4.1.5 Control Rod Guide Tube Split 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.

The original Harris support pins were fabricated from INCONEL Alloy X-750 that was hot rolled, solution treated, and age hardened at various temperatures and times depending on heat, manufacturer, and fabrication date. Support pins made of this material with the associated heat treatments were shown 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 [38, 39]. In response to industry concern, the Alloy X-750 support pins were replaced with Type 316 SS support pins at Harris in the spring of 2006 [39].

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 management effects are below the screening criteria or for which degradation significance is minimal [16]. Section 4.5 of MRP-227, Revision 1-A describes this categorization of the Type 316 SS support pins and defines the Type 316 SS support pins as a No Additional Measures component [11]. 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.

4.1.6 Reactor Vessel Head and Thermal Sleeve Replacement Harris conducted a reactor vessel closure head (RVCH) replacement during the Fall 2019 refueling outage

[43], which included replacement of the thermal sleeves. Thermal sleeve wear in the industry is managed according to PWROG-16003-P [65], as discussed in Section 4.2.8.

4.1.7 Power Uprate Project Harris was originally licensed for a core power rating of 2,775 megawatts thermal (MWt) and has twice sought to increase licensed core power. The original power uprate request sought to increase licensed core power rating to 2900 MWt in conjunction with a steam generator (SG) replacement. Performance of this power uprate was approved by the NRC, and Harris License Amendment No. 107 was issued on October 12, 2001 [40]. The more recent power uprate request sought to increase licensed core power rating to 2,949 MWt through the use of more accurate feedwater flow measurement [41]. Performance of this power uprate was approved by the NRC, and Harris License Amendment No. 139 was issued on May 30, 2012 [42].

WCAP-17209-P, Revision 2, Harris Nuclear Plant Measurement Uncertainty Recapture Power Uprate Engineering Report [41] on the 2012 power uprate summarized the various evaluations and analyses of the power uprate on plant systems, components, and analyses, including the RVI. The NRC staff reviewed the licensees evaluation of the impact of the license amendment request (LAR) on the structural integrity

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-7 WCAP-18710-NP August 2023 Revision 0 assessments for the RVI. The NRC staff determined in the SE of the power uprate application (included in

[42]) that the licensees RVI evaluation considering the effects of the LARs 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 Commissions regulations.
3. The issuance of the amendment 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 Aging Management for Reactor Internals (WCAP-14577)

The WOG topical report WCAP-14577, Revision 1, License Renewal Evaluation: Aging Management for Reactor Internals [14], 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 from the WCAP to develop plant-specific AMPs.

Harris is creating this AMP based on MRP-227, Revision 1-A [11]. WCAP-14577, Revision 1 was an important first step in the development of a generic industry program for managing reactor vessel internals, and was an input and building block for MRP-227, Revision 1-A.

4.2.2 Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A)

MRP-227, Revision 1 was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants, and international committee 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, Revision 1 was accepted by the NRC and reissued as MRP-227, Revision 1-A [11]. The applicability of the MRP-227, Revision 1-A guidelines to Harris is described in detail within Section 6.

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 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:

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-8 WCAP-18710-NP August 2023 Revision 0 Primary Those PWR internals that are highly susceptible to the effects of at least one of the eight 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 Code Section XI [6]) 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 Harris is described in detail within Section 6.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-9 WCAP-18710-NP August 2023 Revision 0 4.2.3 Reactor Internals Acceptance Criteria Methodology and Data Requirements (WCAP-17096-NP-A)

The industry, through various cooperative efforts, has developed a 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 [21], which details acceptance criteria methodology for the MRP-227 Primary and Expansion components. Revision 2 of WCAP-17096-NP was transmitted to the NRC on May 19, 2010 for review. The NRC accepted WCAP-17096-NP, Revision 2 on May 3, 2016, allowing the issuance of WCAP-17096-NP-A [21].

WCAP-17096-NP-A and the guidance within PWROG-17071-NP [22], 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-A. PWROG-17071-NP was issued to the PWROG members within OG-18-61 [44].

The status of WCAP-17096-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 [45] has been sent to the NRC for review and acceptance within OG-19-164 [46]. WCAP-17096, Revision 3 includes the interim guidance from PWROG-17071-NP and updates to the acceptance criteria needed for consistency with MRP-227, Revision 1. It also extends the applicability to include both initial license renewal (60-year operating license) and subsequent license renewal (80-year operating license) based on current industry knowledge.

4.2.4 Reactor Internals Guide Tube Wear (WCAP-17451-P)

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. A technical basis document, WCAP-17451-P, Revision 2, Reactor Internals Guide Tube Wear

- Westinghouse Domestic Fleet Operation Projections [30], 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 2

[30], and includes the modified requirements of interim guidance provided in EPRI letter MRP 2018-007

[47] and PWROG letter OG-18-46 [48] (see Table 4-3 of MRP-227, Revision 1-A). 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 [49].

MRP-227, Revision 1-A designates the Control Rod Guide Tube Assembly: Guide plates (cards) as a Primary component, with no expansion links. Examinations are to be conducted per the requirements of WCAP-17451-P.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-10 WCAP-18710-NP August 2023 Revision 0 WCAP-17451-P Inspection Requirements Harris Unit 1 is a 3-loop plant with a 17x17 Std. guide tube design and is represented in WCAP-17451-P, per [43]. To meet the NEI-03-08 Needed requirements of Section 6 of WCAP-17451-P, Harris will conduct guide card examinations. Table 5-14 of WCAP-17451-P indicates that Harris should perform an initial guide card inspection at 33.0 effective full-power years (EFPY). For plants listed in Table 5-14 of WCAP-17451-P, an alternative initial inspection measurement can be performed during an outage within a time range from the listed EFPY up to four additional EFPY, or an earlier time for measurement or inspection may be selected at the discretion of the utility. Under this guidance, Harris would perform an initial baseline examination measurement no later than 37.0 EFPY. This corresponds to the Spring 2027 outage for Harris (see Table 7-1 of Section 7). Alternatively, Section 5.5.1 of WCAP-17451-P allows utilities to optionally perform video inspections of the guide cards to help screen when guide tubes will need to be measured. This screening may be used to estimate the wear condition of the guide cards and develop an alternate initial wear measurement schedule.

The baseline inspection scope, acceptance criteria, required actions, and reinspection schedule shall be determined according to the guidance within WCAP-17451-P.

Ion-nitride RCCAs at Harris Harris has replaced the 52 Westinghouse 17x17 EP-RCAAs with the AREVA HARMONI design. Critical differences of the AREVA HARMONI design are specified in [28]. The AREVA HARMONI design uses ion nitride cladding for the RCCA.

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 predictions in WCAP-17451-P, Revision 1, as stated in NSAL-17-1 [50]. 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

[51]. WCAP-17451-P, Revision 2 includes updated Needed guidance for addressing accelerated guide card wear and PWROG letter OG-18-276 [52] transmitted this Needed guidance to the PWROG members.

Harris uses ion nitride RCCAs in conjunction with a 17x17 Std. guide tube design. Plants using ion nitride RCCAs in conjunction with 17x17 Std. style guide tubes do not fall under the updated Needed guidance within WCAP-17451-P and OG-18-276. Therefore, per NSAL-17-1, baseline guide card wear measurements can be conducted consistent with the recommendations in WCAP-17451-P.

4.2.5 Core Barrel Operating Experience During spring 2018 inspections, one CE-designed plant identified crack-like surface indications at the core support barrel (CSB) assembly welds, specifically one 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 materials with neutron dose levels high enough for potential IASCC, but the degradation mechanism has

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-11 WCAP-18710-NP August 2023 Revision 0 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 [53].

Potential core barrel cracking was evaluated in PWROG-17084-P [54] to consider the safety significance of the OE in MRP 2018-028 and core barrel cracking in general. The evaluation determined that if the core barrel failed completely as a result of a faulted event and the event itself initiated a reactor shutdown, then the barrel separation would not impede the safe shutdown of the plant. If instead the barrel separation occurred during normal or upset operating conditions, or during a faulted event which does not trigger a reactor shutdown, it could result in an unanalyzed condition. Such a condition was considered low probability but supports the need for the core barrel weld inspection guidance of MRP-227, Revision 1-A.

NEI 03-08 Good Practice interim guidance for MRP-227-A was developed within MRP 2019-009 [55].

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 I&E 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 U.S. PWR units as of February 1, 2020 [55]. Harris falls into Group 2 of MRP-2019-009 and the Good Practice guidance is being evaluated.

During fall 2022 inspections, a 3-loop Westinghouse-designed plant identified crack-like surface indications on the inside diameter of the core barrel upper girth weld (UGW). Specifically, five indications were identified near the bottom of the weld and parallel to it. A supplemental volumetric examination to characterize the indications was subsequently performed. The examination confirmed that the visually identified indications did not expand through the core barrel thickness; however, some of the flaws were greater than 75% through-wall. This OE was communicated to the PWR fleet in MRP 2023-001 [82], and a PWROG focus group was formed to provide a unified approach across the industry to address aging management considerations related to core barrel weld cracking. Duke Energy will participate in this focus group, evaluate the impact of the fall 2022 OE, and address any resulting industry guidance changes at Harris via the NEI 03-08 protocol [7].

4.2.6 Baffle-Former Bolt Degradation (NSAL-16-1)

Recently, a larger-than-expected number of degraded baffle-former bolts were discovered in downflow plants through MRP-227 inspection requirements and reactionary inspections at similar plants. The pattern of these degraded bolts was also concentrated, or clustered, more than anticipated based on the 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 [56]. 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 reinspection techniques and intervals. Interim guidance has been developed and published by the MRP for the Tier 1 plants in MRP 2016-021 [57] and for all of the NSAL-16-1 plants in

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-12 WCAP-18710-NP August 2023 Revision 0 MRP 2017-009 [58]. The interim guidance within MPR 2017-009 has been incorporated into Table 4-3 of MRP-227, Revision 1-A [11].

Harris is a designed upflow plant as discussed in Section 4.1.4, and therefore falls 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, Harris will perform a baseline volumetric (UT) examination no later than 35 EFPY. This corresponds to the Fall 2025 outage for Harris (see Table 7-1 of Section 7).

In accordance with MRP 2018-002 [59], if Harris discovers significant baffle-former bolt clustering as defined in MRP-2018-002, Harris will perform a one-time visual (VT-3) examination of barrel-former bolts within three 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.

4.2.7 Clevis Bearing STELLITE Wear Surface and Clevis Insert Bolts (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. TB-14-5 also reviews the safety implications of the OE and root cause analysis results and provides inspection recommendations for licensees to consider as part of their aging management program to address this OE.

As discussed in EPRI notification letter MRP 2020-11 [60], there have been multiple cases of OE at domestic PWR plants related to degradation of Lower Radial Support Clevis Insert X-750 Bolts, the most recent of which resulted in partial displacement of one insert from its original designed interference fit in the vessel clevis. As delineated in TB-14-5, all Westinghouse-designed PWR plants have been constructed with clevis insert bolting materials with fabrication heat treatments that make them susceptible to PWSCC.

Additional information may be obtained from PWROG-15034-P [61].

Harris conducted VT-3 (Visual) examinations of the clevis insert bolts during the Spring 2018 (27.13 EFPY) refueling outage [63]. No recordable indications were observed.

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 [64]. 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) fell from the RV closure head during an ISI [64]. 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.

PWROG-16003-P [65], 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. A thermal sleeve flange wear inspection was performed at Harris during the Spring 2018 refueling outage in accordance with TB-07-2 using the

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-13 WCAP-18710-NP August 2023 Revision 0 evaluation methodology and acceptance criteria within PWROG-16003-P. This inspection indicated that no thermal sleeve locations exceeded the generic separation acceptance criteria established in PWROG-16003-P. Further plant OE, as described within NSAL-18-1 [66], 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, which can lead to stuck control rods. PWROG-16003-P was updated to Revision 2 [65] to account for the OE reported within NSAL-18-1 and was transmitted to the PWROG members within OG-19-101 [67].

Harris conducted a reactor vessel closure head (RVCH) replacement during the Fall 2019 refueling outage

[43], which included replacement of all thermal sleeves. Harris will use the guidance in PWROG-16003-P for future management of thermal sleeve flange wear. PWROG-16003-P states that there are no recommendations for units that have less than 20 EFPY on their original or replacement reactor vessel head.

Harris falls into this category after replacing the reactor vessel head and will remain in this category until the replacement head accumulates 20 EFPY of operation.

4.2.9 On-Going Industry Programs and NEI 03-08 Guidelines As part of its license renewal, Duke Energy stated that Harris would participate in industry activities associated with RVI-related issues and that the Harris 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 NEI 03-08

[24]. Included work products are MRP-227 [11], MRP-228 [20], WCAP-17451-P [30], 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 Harris is basing the AMP and Inspection Plan for RVI on the scope defined in MRP-227, Revision 1-A, Harris will also implement work products issued under implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [7], including later revisions and interim guidance to the work products listed above. A failure to meet a Needed or 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 [7].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-14 WCAP-18710-NP August 2023 Revision 0 4.3

SUMMARY

This section contains pertinent Harris and industry programs and activities used for the development and implementation of MRP-227, Revision 1-A, the Harris RVI AMP and Inspection Plan, and the Harris 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 Harris Programs, and use of OE, provide reasonable assurance that the reactor internals at Harris will continue to perform their intended functions throughout the period of extended operation.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-1 WCAP-18710-NP August 2023 Revision 0 5

SHEARON HARRIS REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES Based on Duke Energys LRA commitment [1], the Harris RVI AMP and Inspection Plan is credited for aging management of RVI components for the following eight 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 Harris RVI Program and compliance with NUREG-1801 (GALL Report), Section XLM16A, PWR Vessel Internals [9], as updated via SLR-ISG-2021-01 [10] are described in this section.

The Harris RVI Program is aligned to meet the requirements of MRP-227, Revision 1-A [11] in addition to complying with the GALL and SLR-ISG-2021-01. The GALL identifies ten attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the ten elements of the GALL. 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 Harris RVI Program are summarized in the following subsections.

Note that SLR-ISG-2021-01 applies to both the license renewal GALL report, NUREG-1801, and the subsequent license renewal (SLR) GALL report, NUREG-2191 [68]. Harris has not included SLR within the scope of this AMP and Inspection plan; therefore, any updates in SLR-ISG-2021-01 specific to subsequent license renewal or NUREG-2191 are not applicable to this AMP and are not included in the following subsections.

The GALL elements detailed in the following sections make multiple references to MRP-227 (as supplemented) which refers to the gap analysis performed by plants applying for SLR, which should use MRP-227, Revision 1-A as supplemented by a gap analysis. Harris has not included SLR within the scope of this AMP, and therefore does not need to conduct a gap analysis and can directly use the current NRC approved version of MRP-227 (Revision 1-A) in this AMP without a gap analysis. For the purposes of this AMP, references to MRP-227 (as supplemented) within the GALL elements are references to MRP-227, Revision 1-A. The exception to this statement is the reference to MRP-227-A (as supplemented) in GALL element 5, where MRP-227-A [69] should be used instead.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 WCAP-18710-NP August 2023 Revision 0 5.1 GALL ELEMENT 1: SCOPE OF PROGRAM GALL Report AMP Element Description ([9] and [10])

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 (as supplemented), which provides augmented inspection and flaw evaluation guidelines 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. Since these types of AMPs are considered to be living programs by the licensees implementing the programs, the scope of program may also include additional reports, documents, or guidelines recommended for implementation by the EPRI MRP, PWR Owners Group, or industry vendors. This may include (but is not limited to) applicable WCAP or BAW technical/topical reports issued by Westinghouse or B&W, or supplemental guidelines or industry alert letters issued by the EPRI MRP, PWR Owners Group, or industry vendors.

The scope of components includes core support structures, those RVI components that serve an intended license renewal safety function pursuant to criteria in Title 10 of the Code of Federal Regulations (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-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227 (as supplemented).

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-SLR AMP XI.M1, ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.

The program element specifies whether the program is based on an existing program that is consistent with MRP-227, Revision 1-A, with a gap analysis, or the program is based on an acceptable generic report that covers an 80-year service life for the RVI components, such as an approved revision of MRP-227 that considers an operating period of 80 years. If based on MRP-227, Revision 1-A, with a gap analysis, the scope of the program focuses on identification and justification of the following:

a. Components that screen in for additional aging effects or mechanisms when assessed for the 60-80 year operating period.
b. Components that previously screened in for an aging effect or mechanism and the severity of that aging effect or mechanism, could significantly increase for the 60-80 year operating period.
      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 WCAP-18710-NP August 2023 Revision 0

c. Changes to the existing MRP-227, Revision 1-A, program characteristics or criteria, including but not limited to changes in inspection categories, inspection criteria, or primary-to-expansion component criteria and relationships.

Shearon Harris Unit 1 Program Scope The last paragraph of this GALL element, including items a, b, and c, does not apply to this AMP since it provides requirements for SLR, and this AMP has been developed for initial license renewal. Also, note that the third paragraph refers to GALL-SLR Report (NUREG-2191 [68]) AMP XI.M1, which does not apply to this AMP since it provides requirements for SLR. Therefore, GALL Report (NUREG-1801 [9])

AMP XI.M1 should be used since it is applicable to this AMP. Note that Section XI.M1 of the GALL Report is equivalent to section XI.M1 of the GALL-SLR Report.

The Harris 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 thimbles, although not a part of the RVI, are being addressed because of their inclusion in MRP-227, Revision 1-A. The flux thimble tubes extend from an external seal table, through the bottom mounted nozzle penetration, through the lower internals assembly, and finally to the top of the fuel assembly. Additional RVI details are provided in the Harris LRA [1] and UFSAR [23].

The Harris RVI subcomponents that are subject to AMP requirements were provided in Table 3.1.2-1 in the Harris LRA [1]. The table listed each subcomponents intended function(s) and material. The aging effects that require management were identified in the table. A column in the table lists the aging management program that was credited to address the component 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 [2]. Table 3.1.2-1 from the LRA is included in Table B-1.

Duke Energys commitment to implement MRP-227, Revision 1-A necessitates that the aging management strategy in the original LRA be updated. Harris utilizes NUREG-1801, Revision 2 as updated by language in SLR-ISG-2021-01 [10] applicable to initial license renewal, to ensure the aging effects are managed so that the intended function(s) will be maintained consistent with the CLB for the extended period of 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, Revision 1-A, is rooted in the GALL methodology. The guideline applicability of MRP-227, Revision 1-A, Section 2.4 is met by Harris as addressed in Section 6.1 of this AMP. The A/LAI provided by the NRC in the SE on MRP-227, Revision 1-A [11] is met by Harris and demonstration of compliance is addressed in Section 6.2.

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 Harris LRA are based on utilizing these ASME Section XI exams and the augmented inspections derived from MRP-227, Revision 1-A. The

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4 WCAP-18710-NP August 2023 Revision 0 MRP-227, Revision 1-A inspections only augment and do not reduce, alter, or otherwise affect the ASME Section XI requirements.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

5.2 GALL ELEMENT 2: PREVENTATIVE ACTIONS GALL Report AMP Element Description ([9] and [10])

The program relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms [e.g., loss of 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-SLR AMP XI.M2, Water Chemistry.

Shearon Harris Unit 1 Program Preventative Actions The end of this GALL element refers to GALL-SLR Report (NUREG-2191 [68]) AMP XI.M2, which does not apply to this AMP since it provides requirements for SLR. Therefore, GALL Report (NUREG-1801

[9]) AMP XI.M2 should be used since it is applicable to this AMP. Note that Section XI.M2 of the GALL Report is equivalent to section XI.M2 of the GALL-SLR Report.

The Harris 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 Harris Water Chemistry Program [26, 27]. A description and applicability to the Harris RVI Program is provided in the following.

Water Chemistry 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 containments below the system-specific limits. The Harris Water Chemistry Program

[26, 27] is based on the current, approved revisions of the EPRI PWR Primary Water Chemistry Guidelines

[31].

The limits of known detrimental contaminants imposed by the Harris Water Chemistry 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 [9], as updated via SLR-ISG-2021-01 [10].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-5 WCAP-18710-NP August 2023 Revision 0 5.3 GALL ELEMENT 3: PARAMETERS MONITORED OR INSPECTED GALL Report AMP Element Description ([9] and [10])

The program manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking due to SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material due to wear; (c) loss of fracture toughness due to thermal aging and 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 (as supplemented).

Shearon Harris Unit 1 Parameters Monitored or Inspected The GALL element detailed in this section fully applies to Harris. The Harris RVI Program monitors the following aging effects by inspection, in accordance with the guidance of MRP-227, Revision 1-A [11].

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 [20], or the associated Existing Program.

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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-6 WCAP-18710-NP August 2023 Revision 0 (b) 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.

(c) 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 examinations techniques, and by applying applicable reduced fracture toughness properties in supplemental flaw growth or flaw tolerance evaluations that are performed.

(d) Changes in Dimension Changes in dimension are due to void swelling or distortion. Changes in dimension are monitored by visual, volumetric, or physical examination.

(e) 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 Harris 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 as Table C-1 of this document),

the Expansion component inspections from Table 4-6 of MRP-227, Revision 1-A (included as Table C-2 of this document), and the Existing Programs component inspections from Table 4-9 of MRP-227, Revision 1-A (included as Table C-3 of this document). 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 [9], as updated via SLR-ISG-2021-01 [10].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-7 WCAP-18710-NP August 2023 Revision 0 5.4 GALL ELEMENT 4: DETECTION OF AGING EFFECTS GALL Report AMP Element Description ([9] and [10])

The inspection methods are defined and established in Section 4 of MRP-227, Revision 1-A or MRP-227 (as supplemented). 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.

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 (as supplemented) 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 (as supplemented).

In some cases (as defined in MRP-227, Revision 1-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 those established in MRP-227 (as supplemented).

This program element should justify the appropriateness of the inspection methods, sample size criteria, and inspection frequency criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to these criteria from their assessment in MRP-227, Revision 1-A.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-8 WCAP-18710-NP August 2023 Revision 0 Shearon Harris Unit 1 Detection of Aging Effects The last paragraph of this GALL element does not apply to this AMP since it provides requirements for SLR, and this AMP has been developed for initial license renewal.

The Harris RVI Program implements the augmented inspection requirements of Table 4-3, Table 4-6, and Table 4-9 from MRP-227, Revision 1-A [11] for the Primary, Expansion, and Existing Programs 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 Harris RVI Program implements the expansion requirements of Table 5-3 of MRP-227, Revision 1-A (included as Table C-4 of this document). 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 [3, 4, 5, 32].

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: visual, ultrasonic, and physical measurement.

Inspection standards developed by the industry for the application of these techniques for augmented reactor internals inspections are documented in MRP-228 [20].

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 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 Harris 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 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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-9 WCAP-18710-NP August 2023 Revision 0 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. Due to personnel radiation exposure issues, VT-1 and EVT-1 examinations of internals are conducted using remote examination techniques.

Ultrasonic Testing Another method selected for use is volumetric examination. An ultrasonic (UT) examination 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, a 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.

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 [30]. 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. The ET examination, as defined in Harriss response to IEB 88-09 [35] is also considered sufficient to monitor for the applicable aging effect on BMI Thimble Tubes.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-10 WCAP-18710-NP August 2023 Revision 0 Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

5.5 GALL ELEMENT 5: MONITORING AND TRENDING GALL Report AMP Element Description ([9] and [10])

The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227, Revision 1-A and its subsections, or MRP-227 (as supplemented). Component reinspection frequencies for Primary and Expansion category components are defined in specific tables in Section 4 of the MRP-227, Revision 1-A report or in MRP-227 (as supplemented). The examination and re-examinations that are implemented in accordance with MRP-227 (as supplemented), together with the criteria specified in MRP-228, Rev. 3 for inspection standards, 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 component in cases where cracking is detected in an 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 the components that are inaccessible by the applicable inspection technique and the resulting impact on the intended functions(s) of the assembly containing the components.

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 (as supplemented).

Shearon Harris Unit 1 Monitoring and Trending The GALL element detailed in this section fully applies to Harris. The methods for monitoring, recording, evaluating, and trending the results from the RVI inspections under the Harris RVI Program are in accordance with MRP-227, Revision 1-A [11], MRP-228 [20], and ASME Section XI [6]. Note that this element specifies criteria based on MRP-228, Revision 3 but MRP-228, Revision 4 has been used to support Harriss compliance with this element. MRP-228, Revision 4 is the latest version of MRP-228, and it is acceptable to use in place of MRP-228, Revision 3. Likewise, the latest revision of MRP-228 at the time of each exam will be used.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-11 WCAP-18710-NP August 2023 Revision 0 Monitoring is accomplished though implementation of MRP-227, Revision 1-A Primary, Expansion, and Existing Programs inspections (Table 4-3, Table 4-6, and Table 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 of this document, Table C-1, Table C-2, and Table 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 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 Harris 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 Section 6 and Section 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 though updates of the industry guidelines.

Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-12 WCAP-18710-NP August 2023 Revision 0 5.6 GALL ELEMENT 6: ACCEPTANCE CRITERIA GALL Report AMP Element Description ([9] and [10])

Section 5 of MRP-227, Revision 1-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, or MRP-227 (as supplemented) provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion RVI component examination methods. Consistent with the criteria in MRP-227, Revision 1-A, the acceptance criteria for some Expansion category components may be established through performance of a component-specific analysis or component replacements, particularly if the components are inaccessible for inspection or the industry has yet to develop adequate inspection methods for the components. 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 appliable, the program establishes acceptance criteria for any physical measurement monitoring methods that are credited for aging management of particular RVI components.

This program element should justify the appropriateness of the acceptance criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to acceptance criteria based on the gap analysis.

Shearon Harris Unit 1 Acceptance Criteria The last paragraph of this GALL element does not apply to this AMP since it provides requirements for SLR, and this AMP has been developed for initial license renewal.

The Harris RVI program acceptance criteria for the Westinghouse-designed Primary and Expansion component examinations are consistent with Section 5 of MRP-227, Revision 1-A [11]. For the Westinghouse-designed Existing Programs components, the acceptance criteria are described within the applicable program documents. The Harris 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

[30].

Examination acceptance and expansion criteria for the MRP-227, Revision 1-A inspections are provided in Table C-4 of this document. The Existing Programs components in Table C-3 will continue to be examined in accordance with the credited Existing Programs requirements. Augmented inspections, as defined by the MRP-227, Revision 1-A requirements included in Table C-1 and Table C-2, that result in relevant indications will be managed by the Corrective Action Program [70, 71] 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 Section 6 and Section 7.5 of MRP-227, Revision 1-A.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-13 WCAP-18710-NP August 2023 Revision 0 Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

5.7 GALL ELEMENT 7: CORRECTIVE ACTIONS GALL Report AMP Element Description ([9] and [10])

Results that do not meet the acceptance criteria are addressed in the applicants corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, Corrective Action, of 10 CFR Part 50, Appendix B. Appendix A of the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

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 (as supplemented), plus the 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.

Shearon Harris Unit 1 Corrective Actions The first paragraph of this GALL element refers to Appendix A of the GALL-SLR Report (NUREG-2191

[68]), which does not apply to this AMP since it provides requirements for SLR. Therefore, Appendix A of the GALL Report (NUREG-1801 [9]) should be used since it is applicable to this AMP. Note that Appendix A of the GALL Report is equivalent to Appendix A of the GALL-SLR Report.

Relevant indications discovered during examinations of the Harris RVI will be managed through the Corrective Action Program [70, 71] and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation actions, or analytical evaluations. 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 [11] provide the basis for what relevant conditions adverse to quality must be addressed by the Corrective Action Program.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-14 WCAP-18710-NP August 2023 Revision 0 The ASME Section XI Program [3, 4, 5, 32] establishes the Harris 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 [9], as updated via SLR-ISG-2021-01 [10].

5.8 GALL ELEMENT 8: CONFIRMATION PROCESS GALL Report AMP Element Description ([9] and [10])

The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, Correction Action, 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

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 Section 7 of MRP-227, Revision 1-A, in conjunction with NEI 03-08 and other guidance documents, reports, or guidelines 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.

Shearon Harris Unit 1 Confirmation Process The first paragraph of this GALL element refers to Appendix A of the GALL-SLR Report (NUREG-2191

[68]), which does not apply to this AMP since it provides requirements for SLR. Therefore, Appendix A of the GALL Report (NUREG-1801 [9]) should be used since it is applicable to this AMP. Note that Appendix A of the GALL Report is equivalent to Appendix A of the GALL-SLR Report.

Harris has established a 10 CFR 50, Appendix B Program [72] 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. Harris implements work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [7]. 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 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 [11], 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 Harris RVI.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-15 WCAP-18710-NP August 2023 Revision 0 Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

5.9 GALL ELEMENT 9: ADMINISTRATIVE CONTROLS GALL Report AMP Element Description ([9] and [10])

Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging.

Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

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 basis defined in Section 7 of MRP-227, Revision 1-A, found acceptable as documented in the staffs safety evaluation dated April 25, 2019, provides the basis for implementing the program in accordance with NEI 03-08.

Administrative activities for keeping the program implementation procedures up to date with the various industry reports within the scope of the AMP (e.g., MRP-227, Revision 1-A) fall within the scope of this Administrative Controls program element.

Shearon Harris Unit 1 Administrative Controls The first paragraph of this GALL element refers to Appendix A of the GALL-SLR Report (NUREG-2191

[68]), which does not apply to this AMP since it provides requirements for SLR. Therefore, Appendix A of the GALL Report (NUREG-1801 [9]) should be used since it is applicable to this AMP. Note that Appendix A of the GALL Report is equivalent to Appendix A of the GALL-SLR Report.

Harris has an established 10 CFR 50, Appendix B Program [72] that addresses the elements of corrective actions, confirmation process, and administrative controls. The Harris program includes safety and non-safety related structures, systems, and components. QA procedures, review and approval processes, and administrative controls are implemented with the requirements of 10 CFR 50, Appendix B.

Harris implements work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures [7]. These procedures also implement activities to update the RVI program with the issuance of industry documents that can impact the program. 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 of NEI 03-08 and Duke Energy administrative procedures, including notification to the NRC for information only.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-16 WCAP-18710-NP August 2023 Revision 0 Conclusion This element is consistent with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [9], as updated via SLR-ISG-2021-01 [10].

5.10 GALL ELEMENT 10: OPERATING EXPERIENCE GALL Report AMP Element Description ([9] and [10])

The review and assessment of relevant operating experience (OE) for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227, Revision 1-A. Consistent with MRP-227, Revision 1-A, the reporting of inspection results and OE 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 OE including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

Shearon Harris Unit 1 Operating Experience The last paragraph of this GALL element refers to Appendix B of the GALL-SLR Report (NUREG-2191

[68]), which does not apply to this AMP since it provides requirements for SLR. Therefore, Appendix B of the GALL Report, as documented in LR-ISG-2011-05 [73], should be used since it is applicable to this AMP. Note that Appendix B of the GALL Report is equivalent to Appendix B of the GALL-SLR Report.

Extensive industry and Harris OE have been reviewed during the development of the RVI AMP. The experience review includes NRC Information Notices 84-18, Stress Corrosion Cracking in PWR Systems

[74] and 98-11, Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants [75]. Most of the industry OE review 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, core barrel, and control rod guide tube guide cards has also been reviewed.

Early plant OE related to hot functional testing and reactor internals is documented in plant historical records. Inspections performed as part of the ASME Section XI program [3, 4, 5, 32] have been conducted as designated by existing commitments and are expected to discover general internals structure degradation.

To date, little degradation has been observed industry wide.

A review of industry and plant-specific experience with RVI reveals that the U.S. Industry, including Duke Energy and Harris, 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 Harris in the Spring of 2006. The replacement pins included a material upgrade to Type 316 SS in support of managing aging in the component. Duke Energy will address industry guidance on accelerated guide card wear, as discussed in Section 4.2.4; on the core barrel welds, as

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-17 WCAP-18710-NP August 2023 Revision 0 discussed in Section 4.2.5; on the baffle-former bolts, as discussed in Section 4.2.6; and on the clevis bearing STELLITE wear surface and clevis insert bolts as discussed in Section 4.2.7.

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 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 PWR internals 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 Harris operating experience procedure [76], 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 [9], as updated via LR-ISG-2011-05 [73].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-1 WCAP-18710-NP August 2023 Revision 0 6

MRP-227 SAFETY EVALUATION CONDITIONS AND ACTION ITEMS The NRC accepted MRP-227, Revision 1-A [11] to the extent delineated within the NRC SE on MRP-227, Revision 1, but requested that utilities who reference MRP-227, Revision 1-A address the items described within [77]. These items were addressed by the industry within MRP 2020-012 [78]. The NRC issued an SE [79] accepting MRP 2020-012 as an addendum or errata of MRP-227, Revision 1-A and concluded that MRP 2020-012 achieved the following objectives with regard to the items raised in [77]:

1. Clarified the EPRI MRPs reasons for a few component nomenclature differences for some component descriptions that were included in the various tables of the MRP-227, Rev. 1-A report, in which the staff sought additional clarifications from the EPRI MRP. The staff found the EPRI MRPs explanations provided an acceptable basis for the differences in the component nomenclatures between tables in the report.
2. Clarified that changes in specific footnote designations for specific TR table footnotes specified in the staffs February 20, 2020, were administrative in nature.
3. Clarified that no further edits of the MRP-227, Rev. 1-A are necessary.

Out of five items listed within Table 1 of MRP-2020-012, only items 5a, 5b, 5d, and 5e pertain to Harris.

As concluded in MRP-2020-012, the note numbering changes described in these items were confirmed to be administrative, editorial adjustments and do not require further action beyond referencing the explanations provided in Table 1. The NRC SE on MRP 2020-012 concluded that items raised in [77] have been sufficiently addressed by the industry in MRP 2020-012. Therefore, no additional explanation is required in this report to satisfy [77].

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 by Harris.

6.1 MRP-227, REVISION 1-A GUIDELINE APPLICABILITY The MRP-227, Revision 1-A [11] 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 assessments and the resultant supporting aging management strategies in MRP-232 [18]

provide the basis for the MRP-227, Revision 1-A guidelines for Westinghouse and Combustion Engineering designed plants. These evaluations were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter. This section describes the guidelines of applicability of MRP-227, Revision 1-A and Harriss compliance.

The applicability of MRP-227, Revision 1-A to Harris requires compliance with the following MRP-227, Revision 1-A general assumptions:

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 WCAP-18710-NP August 2023 Revision 0 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.

Harris Applicability Harris operated for approximately 2.00 effective full-power years (EFPY) with fresh fuel assemblies at peripheral locations (high-leakage core loading pattern) [81]. Harris transitioned to a low-leakage fuel management strategy beginning with Cycle 3. Since that time, the plant has implemented low-leakage core designs in every cycle except Cycle 11. During Cycle 11, Harris operated with a high-leakage core loading pattern, corresponding to an additional 1.28 EFPY [43].

Combined, Harris has operated for 3.28 EFPY with a high-leakage core loading pattern. The current low-leakage core design philosophy is anticipated for the extended plant operating license

[81]. By operating with a high leakage core design for less than 30 years of operation, Harris 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 Harris core geometry and operating characteristics to the applicability guidelines for Westinghouse-designed reactors specified in Appendix B was performed for Harris within [81].

This comparison concluded that Harris has not utilized atypical fuel design or fuel management, including power changes/uprates that have occurred over its operating lifetime, that could make the assumptions of MRP-227, Revision 1-A regarding core loading/core design non-representative.

Harris has operated for less than 30 years with a high-leakage core loading pattern and operates with fuel design and management that meets the requirements of MRP-227, Revision 1-A, Appendix B. Therefore, Harris 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.

Harris Applicability Harris currently operates as base-loaded power plant and has operated under base load conditions for the majority of the life of the plant [43]. Therefore, Harris satisfies this assumption in MRP-227, Revision 1-A.

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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 WCAP-18710-NP August 2023 Revision 0 Harris Applicability Plant modifications impacting the Harris reactor internals made over the lifetime of the plant are those specifically directed by Westinghouse, the original equipment manufacturer (OEM) [43], and include the control rod guide tube (CRGT) support (split) pin replacement project [39], two power uprate projects in [40] and [41, 42], and a reactor vessel head replacement [43]. Therefore, Harris 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.

Harris Applicability The typical Westinghouse PWR internals components are listed within Table 4-4 of MRP-191, Revision 1 [16]. The Harris internals components and materials were identified in the Harris LRA

[1] and confirmed by referencing the applicable Harris component drawings [43]. The internals components and materials were then compared to those in Table 4-4 of MRP-191, Revision 1 in

[43]. The comparison of typical Westinghouse PWR internals components to the Harris PWR internals components is described in the following.

a. Confirmation that no additional items are identified by this comparison:

The components identified within Table 2.3.1-1 of the Harris LRA were compared to those in Table 4-4 of MRP-191, Revision 1, within [43]. This review did not identify any components that were not specifically included within Table 4-4 of MRP-191, Revision 1.

It is noted that the Harris LRA identifies the lower internals; diffuser plate as a component at Harris. However, the review of the applicable component drawings indicated that Harris does not have a diffuser plate. In correspondence attached to [43], Duke Energy confirmed that there is no diffuser plate at Harris. As such, this component is not considered further.

b. Confirmation that the materials identified for Harris are consistent with those materials identified in Table 4-4 of MRP-191:

The component materials identified within the applicable drawings for the components in Table 2.3.1-1 of the Harris LRA were compared to those of MRP-191, Revision 1 within

[43]. All of the materials for Harris are identical or equivalent to those identified in MRP-191, Revision 1, Table 4-4 for Westinghouse-designed plants. 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. At Harris, this includes components fabricated from Type 304L SS instead of Type 304 SS.

Types 304 SS and 304L SS fall under the austenitic stainless-steel category, and there are no differences in the screening criteria for the materials [15, 43]. With no changes to the susceptibility or degradation mechanisms of concern, the FMECA and functionality analysis are acceptable for these equivalent materials.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 WCAP-18710-NP August 2023 Revision 0 It is noted that the BMI column assemblies - BMI column cruciforms component material is listed in Table 4-4 of MRP-191, Revision 1 as Type 304 SS; however, according to the Harris component drawings, the material at Harris is CF8 [43]. Upon further investigation, it was discovered that Table 5-1, Table 6-5, and Table 7-2 of MRP-191, Revision 1 list CF8 as an applicable material for this component. This indicates that CF8 was included as a material for the BMI column cruciforms during the Expert Panel review and screening, categorization, and ranking for MRP-191, Revision 1. Therefore, the absence of this material from Table 4-4 is judged to be an oversight, and this component material is considered to be consistent with MRP-191, Revision 1.

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

Harris internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication [43].

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.

Harris Applicability Harris does not have any major plant-specific differences from the inputs to the FMECA process.

Inputs for the Harris RVI have already been confirmed for previous assumptions regarding core loading pattern history, operating history, design changes, and components/materials. The following describes a comparison of the remaining FMECA inputs to the Harris RVI.

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

The Harris reactor coolant system operates at Thot and Tcold of 601.4°F and 548.4°F, respectively [43]. These are the design basis parameters for the plant. Therefore, Harris satisfies the assumptions in MRP-227, Revision 1-A regarding temperature operational parameters.

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

Stress values are not readily available for the Harris components; however, because the plant design was maintained per Westinghouse directives over the lifetime of the plant [43],

the Harris values are represented by the stresses assumed in MRP-191, Revision 1 and MRP-232, Revision 1. Therefore, the applicability of the general FMECA is confirmed, and the stress values at Harris are consistent with the assumptions in MRP-227, Revision 1-A.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-5 WCAP-18710-NP August 2023 Revision 0 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.

Harris Applicability No modifications have been made to the Harris RVI after calendar year 2007, except for the thermal sleeve replacements, which were conducted as part of the RVCH replacement during the Fall 2019 refueling outage [43]. Thermal sleeve flange wear is not addressed in MRP-227, Revision 1-A. In the response to NRC RAI 28 on MRP-227, Revision 1-A, EPRI indicated that thermal sleeve flange degradation would be addressed in Revision 2 of MRP-227. Guidance regarding thermal sleeve wear has been issued by the industry and documented in PWROG-16003-P, Revision 2 [65]. By replacing the thermal sleeves, Harris has taken a proactive approach to addressing this industry wear issue, and there are no concerns that this change will impact the strategies contained in MRP-227, Revision 1-A. The design has been maintained over the lifetime of the plant as specified by the OEM. Therefore, Harris 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.

Harris Applicability Harris is compliant with the applicable requirements in Table 4-9 of MRP-227, Revision 1-A. This is detailed in the Harris-applicable Duke administrative procedure documents for ASME Section XI in [3] and [4], and the Harris flux thimble tube program [29]. Harris has a number of programs and activities that support the aging management of the RVI, such as the Harris Water Chemistry Program in [26] and [27], and the Duke Energy reactor internals program [25].

Harris replaced the INCONEL Alloy X-750 support pins with cold-worked Type 316 SS support pins [12] [39]. As described in Section 4.5 of MRP-227, Revision 1-A, the degradation of support pins is not a safety issue but is an asset management concern. The Type 316 SS support pins are considered a Category A component and are classified as a no additional measures component.

Therefore, the Type 316 SS support pins do not require a plant-specific aging management program and are not included within Table 4-9 of MRP-227, Revision 1-A. There are no additional inspections required for the Harris Type 316 SS support pins.

Summary This section has demonstrated that Harris meets the requirements for application of MRP-227, Revision 1-A as a strategy for managing age-related material degradation in reactor internals components.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-6 WCAP-18710-NP August 2023 Revision 0 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.

Harris Compliance:

Harris has not yet performed baffle-former bolt inspections. Harris is categorized as Tier 4 within NSAL-16-1 [56]. As directed in MRP-227, Revision 1-A (in accordance with MRP 2017-009 [58] and MRP 2017-010 [80]), Harris will perform the baseline UT inspection of baffle-former bolts no later than 35 EFPY.

Harris 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 Harris will document and disposition the findings in the Harris corrective action program [71]. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 (i.e., 5% of baffle-former bolts with UT or visual indications or clustering as defined in NSAL-16-1) is observed, Harris will follow the interim guidance within MRP 2017-009 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 reinspection 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 Harris 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.

Conclusion Harris 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 Harris.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-1 WCAP-18710-NP August 2023 Revision 0 7

INSPECTION PLAN AND IMPLEMENTATION SCHEDULE The components identified in Table 7-1 cover the Primary component inspection scope for Shearon Harris Unit 1 and are based on Table 4-3 of MRP-227, Revision 1-A [11] 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 provides the due dates for the listed Primary components; these represent the latest opportunity for Harris to perform the initial examinations. Harris 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, which begins at midnight on October 25, 2026 [1]. Additionally, the Harris refueling outages corresponding to the inspection due dates that are based on EFPY in Table 7-1 are conservative estimates based on 18-month fuel cycles and an assumed 1.5 EFPY per cycle 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 Harris. Table C-2 of this document provides the Expansion components from MRP-227, Revision 1-A, the applicability to Harris, 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 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 mechanisms(s) being managed, the existing programs 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 Programs requirements.

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

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-2 WCAP-18710-NP August 2023 Revision 0 Table 7-1 Shearon Harris Unit 1 Primary Component Inspection Plan RVI Component (Note 1)

Examination Method Examination Frequency Harris Due Date (Refueling Outage)

Projected EFPY (Note 2)

W1. Control Rod Guide Tube Assembly Guide plate (cards)

Per the Requirements of WCAP-17451-P Per the requirements of WCAP-17451-P, including subsequent examinations. (Notes 3 and 4)

RFO-27 Spring 2027 35.9 W2. Control Rod Guide Tube Assembly Lower flange welds Enhanced visual (EVT-1) examination No later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

RFO-28 Fall 2028 37.4 W3. Core Barrel Assembly Upper flange weld (UFW)

Enhanced visual (EVT-1) examination No later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

RFO-28 Fall 2028 37.4 W4. Core Barrel Assembly Lower girth weld (LGW)

Enhanced visual (EVT-1) examination No later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

RFO-28 Fall 2028 37.4 W6. Baffle Former Assembly Baffle-former bolts Volumetric (UT) examination Dependent on plant design. Subsequent examination is dependent on the plant design and the results of the baseline inspection. (Notes 5 and 6)

RFO-26 Fall 2025 34.4 W7. Baffle Former Assembly Assembly Visual (VT-3) examination Baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

RFO-29 Spring 2030 38.9

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-3 WCAP-18710-NP August 2023 Revision 0 Notes:

(1) 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 Harris. See Table C-1 of this document for further detail.

(2) Projected EFPY values were calculated based on the total EFPD at the end of cycle R23 (29.9 EFPY) with 1.5 EFPY per cycle (1 EFPY per year) conservatively assumed for the future.

(3) Due to the timing of the associated NRC reviews of industry documents and issuance of WCAP-17451-P, Revision 2 [30], MRP-227, Revision 1-A states utilities should Use WCAP-17451-P, Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [47] and PWROG letter OG-18-46 dated 2/20/2018 [48].

However, Harris will follow the guidance in WCAP-17541-P, Revision 2 for management of guide card and lower guide tube continuous guidance wear at Shearon Harris Unit 1.

(4) Section 5.5.1 of WCAP-17451-P allows utilities to optionally perform video inspections of the guide cards to help screen when guide tubes will need to be measured. This screening may be used to estimate the wear condition of the guide cards and develop an alternate initial wear measurement schedule.

(5) Shearon Harris Unit 1 is a Tier 4 plant in NSAL-16-1 [56]. In accordance with MRP-227, Revision 1-A, Harris will perform baseline UT examination no later than 35 EFPY.

(6) 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 [58] (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 [57] and MRP 2017-009) 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 [19]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

  • Clustering is defined per NSAL-16-1, Revision 1 [56] 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.
      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-1 WCAP-18710-NP August 2023 Revision 0 8

SUMMARY

AND CONCLUSIONS This report documents and provides a description of the Harris RVI AMP and Inspection Plan and how it relates to the Harris RVI Program for management of aging effects through the period of extended operation.

This Harris RVI AMP and Inspection Plan is based on MRP-227, Revision 1-A [11]. Section 5 demonstrates that the Harris RVI Program will meet the intent of the 10 AMP elements described in Chapter XI, AMP XI.M16A of NUREG-1801, Revision 2 [9] as modified by LR-ISG-2011-05 [73] and SLR-ISG-2021-01

[10]. Section 6 of this document confirms the applicability of MRP-227, Revision 1-A to Harris and addresses the applicant/licensee plant-specific action item from MRP-227, Revision 1-A.

The Harris 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 Harris UFSAR to the inspection guidelines provided in MRP-227, as approved by the NRC. This Harris RVI AMP and Inspection Plan, as documented herein, is based on MRP-227, Revision 1-A. Once the Harris RVI AMP and Inspection Plan is approved by the NRC, the Harris UFSAR will be updated as required.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-1 WCAP-18710-NP August 2023 Revision 0 9

REFERENCES

1.

Duke Energy License Renewal Application, Harris Nuclear Plant License Renewal Application, November 2006.

2.

NUREG-1916, Safety Evaluation Report Related to the License Renewal of Shearon Harris Nuclear Power Plant, Unit 1, U.S. Nuclear Regulatory Commission, Docket No. 50-400, August 2008. (NRC: ML082340985)

3.

AD-EG-ALL-1702, Most Recent Revision, ASME Section XI Inservice Inspection Program Administration. (Duke Energy Document)

4.

AD-EG-ALL-1720, Most Recent Revision, Inservice Testing (IST) Program Implementation.

(Duke Energy Document)

5.

PLP-605, Most Recent Revision, ASME Boiler and Pressure Vessel Code Section XI Repair/Replacement Activity Program, (Duke Energy Document)

6.

ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition with the 2008 Addenda.

7.

AD-EG-ALL-1912, Most Recent Revision, Materials Degradation Management Program Implementation. (Duke Energy Document)

8.

10 CFR 54, Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, U.S. Nuclear Regulatory Commission.

9.

NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report - Final Report, U.S.

Nuclear Regulatory Commission, December 2010. (NRC: ML103490041)

10.

SLR-ISG-2021-01, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors Interim Staff Guidance U.S. Nuclear Regulatory Commission, January 2021. (NRC: ML20217L203)

11.

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

12.

WCAP-17101-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Harris Nuclear Plant, November 2009. (Westinghouse Report)

13.

NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants - Final Report, U.S. Nuclear Regulatory Commission, December 2010.

(NRC: ML103490036)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2 WCAP-18710-NP August 2023 Revision 0

14.

WCAP-14577, Revision 1-A, License Renewal Evaluation: Aging Management for Reactor Internals, March 2001. (Westinghouse Report)

a. WCAP-14577-R1-A-ASMT-2, Short Form Revision for License Renewal Evaluation: Aging Management for Reactor Internals, August 2017. (Westinghouse Assessment Record)
15.

Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1). EPRI, Palo Alto, CA: 2017. 3002010268.

16.

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.

17.

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

1021026.

18.

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

1021029.

19.

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. (U.S. Nuclear Regulatory Commission Letter) (NRC: ML19081A001)

20.

MRP-228, Revision 4, Materials Reliability Program: Inspection Standard for Pressurized Water Reactor Internals - 2020 Update. EPRI, Palo Alto, CA: 2020. 3002018245.

21.

WCAP-17096-NP-A, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements, August 2016. (Westinghouse Report) (NRC: ML16279A320)

22.

PWROG-17071-NP, Revision 0, WCAP-17096-NP-A Interim Guidance, March 2018. (PWR Owners Group Report) (NRC: ML18204A166)

23.

Shearon Harris Nuclear Power Plant Updated Final Safety Analysis Report, through Amendment

63.
24.

NEI 03-08, Revision 4, Guideline for the Management of Materials Issues, Nuclear Energy Institute, Washington, DC, October 2020. (NRC: ML20315A536)

25.

AD-EG-ALL-1911, Most Recent Revision, Reactor Internals Program Implementation. (Duke Energy Document)

26.

AD-CP-PWR-0001, Most Recent Revision, Primary Water Chemistry Program. (Duke Energy Document)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-3 WCAP-18710-NP August 2023 Revision 0

27.

CSD-CP-HNP-0001, Most Recent Revision, Harris Primary Chemistry Strategic Plan. (Duke Energy Document)

28.

EC 69253, Revision 0, Engineering Change 69253R0 Section B, Date Unknown. Duke Retrievable as EC 269253, Replace All 52 Rod Cluster Control Assemblies (RCCAs) (A.K.A Control Rods), July 27, 2009. (Duke Energy Document)

29.

EPT-114, Most Recent Revision, Eddy Current Testing Requirements for the Incore Instrumentation Thimbles. (Duke Energy Document)

30.

WCAP-17451-P, Revision 2, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections, November 2018. (Westinghouse Report)

31.

Pressurized Water Reactor Primary Water Chemistry Guidelines: Volume 1, Revision 7. EPRI, Palo Alto, CA; 2014, 3002000505.

32.

AD-EG-ALL-1704, Most Recent Revision, Augmented Inservice Inspection Program Administration. (Duke Energy Document)

33.

TB-14-5, Revision 0, Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation, August 25, 2014. (Westinghouse Technical Bulletin)

34.

TB-16-4, Revision 0, Fuel Alignment Pin Malcomized Surface Degradation, August 15, 2016.

(Westinghouse Technical Bulletin)

35.

NRC Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors, July 26, 1988. (U.S.

Nuclear Regulatory Commission Bulletin)

36.

WCAP-12866, Revision 0, Bottom Mounted Instrumentation Flux Thimble Wear, January 1991.

(Westinghouse Report)

37.

CQLM-10559, Transfer of Reactor Internals, August 30, 1982. (Westinghouse Field Change Notice)

38.

WCAP-15028-NP, Revision 1, Guide Tube Cold-Worked 316 Replacement Support Pin Development Program, June 2011. (Westinghouse Report)

39.

WCAP-16470-P, Revision 0, Shearon Harris Unit 1 Replacement CW 316 SS Support Pin Design Qualification Report, October 2005. (Westinghouse Report)

40.

Sharon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment RE: Steam Generator Replacement and Power Uprate (TAC Nos. MB0199 and MB0782), October 12, 2001. (U.S.

Nuclear Regulatory Commission Letter) (NRC: ML012830516)

41.

WCAP-17209-P, Revision 2, Harris Nuclear Plant Measurement Uncertainty Recapture Power Uprate Engineering Report, May 2012. (Westinghouse Report)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-4 WCAP-18710-NP August 2023 Revision 0

42.

Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment RE: Measurement Uncertainty Recapture Power Uprate (TAC No. ME6169), May 30, 2012. (U.S. Nuclear Regulatory Commission Letter) (NRC: ML11356A096)

43.

CQL-RV010-CN-CA-000003, Revision 0, Harris Unit 1 Reactor Internals MRP-227, Revision 1-A Guideline Applicability and Licensee Action Item 1, June 29, 2023. (Westinghouse Calculation Note)

44.

OG-18-61, Revision 0, Transmittal of Final Report, PWROG-17071-NP, Revision 0, WCAP-17096-NP-A Interim Guidance (PA-MSC-1567), March 15, 2018. (PWR Owners Group Letter)

45.

WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, July 2019. (Westinghouse Report)

46.

OG-19-164, Revision 0, Transmittal of WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, PA-MSC-15-67, July 31, 2019. (PWR Owners Group Letter)

47.

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

(Materials Reliability Program Letter)

48.

OG-18-46, Revision 0, 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, 2018. (PWR Owners Group Letter)

49.

OG-19-197, Revision 0, 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), September 13, 2019. (PWR Owners Group Letter)

50.

NSAL-17-1, Revision 0, Guide Tube Guide Card Wear Attributed to Ion Nitride Rod Cluster Control Assembly, January 16, 2017. (Westinghouse Nuclear Safety Advisory Letter)

51.

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

52.

OG-18-276, Revision 0, 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. (PWR Owners Group Letter)

53.

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

54.

PWROG-17084-P, Revision 1, Core Barrel Girth and Flange Weld Aging Management Strategy -

Probability and Consequences of Failure, March 2019. (PWR Owners Group Report)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-5 WCAP-18710-NP August 2023 Revision 0

55.

MRP 2019-009, 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, July 17, 2019. (Materials Reliability Program Letter) (NRC: ML19249B102)

56.

NSAL-16-1, Revision 1, Baffle-Former Bolts, August 1, 2016. (Westinghouse Nuclear Safety Advisory Letter) (NRC: ML16222A513)

57.

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.

(Materials Reliability Program Letter) (NRC: ML16211A054)

58.

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. (Materials Reliability Program Letter) (NRC: ML17087A106)

59.

MRP 2018-002, Transmittal of NEI-09-08 Needed Interim Guidance Regarding MRP-227-A and MRP-22, Revision 1 Baffle-Former Bolt Expansion Inspection Requirements for PWR Plants, January 17, 2018. (Materials Reliability Program Letter)

60.

MRP 2020-11, Notification of Recent Operating Experience (OE) Related to Displaced PWR Clevis Insert Assembly," April 28, 2020. (Materials Reliability Program Letter)

61.

PWROG-15034-P, Revision 0, Clevis Bolt Fabrication and Inspection Assessment, January 2016. (PWR Owners Group Report)

62.

PWROG-19003-P, Revision 1, Clevis Insert / Radial Key Wear Assessment, May 2021. (PWR Owners Group Report)

63.

MRP 2018-025, 2018 Biennial Report of Recent MRP-227-A Reactor Internals Inspection Results (Project 694), July 19, 2018. (Materials Reliability Program Letter) (NRC: ML18204A161)

64.

TB-07-2, Revision 3, Reactor Vessel Head Adapter Thermal Sleeve Wear, December 7, 2015.

(Westinghouse Technical Bulletin)

65.

PWROG-16003-P, Revision 2, Evaluation of Potential Thermal Sleeve Flange Wear, May 2019.

(PWR Owners Group Report)

66.

NSAL-18-1, Revision 0, Thermal Sleeve Flange Wear Leads to Stuck Control Rod, July 9, 2018.

(Westinghouse Nuclear Safety Advisory Letter) (NRC: ML18198A275)

67.

OG-19-101, Revision 0, Transmittal of Final PWROG Report PWROG-16003-P, Revision 2, Evaluation of Potential Thermal Sleeve Flange Wear, (PA-MSC-1654R1), May 13, 2019. (PWR Owners Group Letter)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-6 WCAP-18710-NP August 2023 Revision 0

68.

NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR)

Report, U.S. Nuclear Regulatory Commission, July 2017. (NRC: ML17187A031, ML17187A204)

69.

MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. EPRI, Palo Alto, CA: 2011. 1022863.

70.

AD-PI-ALL-0100, Most Recent Revision, Corrective Action Program. (Duke Energy Document)

71.

AD-PI-ALL-0200, Most Recent Revision, Performance Trending. (Duke Energy Document)

72.

DUKE-QAPD-001, Most Recent Revision, Quality Assurance Program Description Operating Fleet. (Duke Energy Corporation Topical Report)

73.

LR-ISG-2011-05, Final License Renewal Interim Staff Guidance LR-ISG-2011-05: Ongoing Review of Operating Experience, U.S. Nuclear Regulatory Commission, March 2012. (NRC:

ML12044A215)

74.

Information Notice 84-18, Stress Corrosion Cracking in Pressurized Water Reactor Systems, March 7, 1984. (U.S. Nuclear Regulatory Commission Information Notice)

75.

Information Notice 98-11, Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants, March 25, 1998. (U.S. Nuclear Regulatory Commission Information Notice)

76.

AD-PI-ALL-0400, Most Recent Revision, Operating Experience Program. (Duke Energy Document)

77.

U.S. Nuclear Regulatory Commission Verification Letter for Electronic 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. (U. S.

Nuclear Regulatory Commission Letter) (NRC: ML20006D152)

78.

MRP 2020-012, Responses to Nuclear Regulatory Commission Staff Comments on MRP-227, Revision 1-A, May 4, 2020. (Materials Reliability Program Letter)

79.

MRP-227 Rev 1-A Generic Resolution Final SE, May 28, 2020. (U.S. Nuclear Regulatory Commission Topical Report Safety Evaluation) (NRC: ML20141L315)

80.

MRP 2017-010, Summary White Paper of the Baffle-Former Bolt Prediction Results Provided by Structural Integrity Associates, AREVA, and Westinghouse, March 17, 2017. (Materials Reliability Program Letter) (NRC: ML17222A169)

81.

CQL-REAC-TM-AA-000001, Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris, October 7, 2022. (Westinghouse Letter)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-7 WCAP-18710-NP August 2023 Revision 0

82.

MRP 2023-001, Notification of Recent PWR Core Barrel Cracking Operating Experience and Recommended PWR Plant Actions, February 2, 2023. (Materials Reliability Program Letter)

A recent revision transmitted by Duke in response to the AMP input request is electronically attached to

[43] within the DCP-21-30_Responses.zip archive.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-1 WCAP-18710-NP August 2023 Revision 0 APPENDIX A ILLUSTRATIONS Figure A-1: Illustration of Harris Internals

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-2 WCAP-18710-NP August 2023 Revision 0 Figure A-2: Harris Control Rod Guide Cards

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-3 WCAP-18710-NP August 2023 Revision 0 Figure A-3: Harris Control Rod Guide Tube Assembly

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-4 WCAP-18710-NP August 2023 Revision 0 Figure A-4: Harris Core Barrel Welds (Neutron Panels Removed for Clarity)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-5 WCAP-18710-NP August 2023 Revision 0 Figure A-5: Harris Baffle Plates (Octant-Symmetric)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-6 WCAP-18710-NP August 2023 Revision 0 Figure A-6: Harris Core Baffle/Barrel Structure

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-7 WCAP-18710-NP August 2023 Revision 0 Figure A-7: Harris Baffle-Former Structure

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-8 WCAP-18710-NP August 2023 Revision 0 Figure A-8: Harris Lower Core Support Structure

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-9 WCAP-18710-NP August 2023 Revision 0 Figure A-9: Harris Lower Core Support Structure - Core Support Forging Cross-Section

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-10 WCAP-18710-NP August 2023 Revision 0 Figure A-10: Harris Core Support Column

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-11 WCAP-18710-NP August 2023 Revision 0 Figure A-11: Harris Bottom-Mounted Instrumentation (BMI) Column Design

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-12 WCAP-18710-NP August 2023 Revision 0 Figure A-12: Harris Upper Internals Assembly

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-1 WCAP-18710-NP August 2023 Revision 0 APPENDIX B SHEARON HARRIS UNIT 1 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

The content and numerical identifiers in Table B-1 are extracted from Table 3.1.2-1, Reactor Vessel, Internals, and Reactor Coolant System -

Summary of Aging management Evaluation - Reactor Vessel and Internals, of the Harris LRA [1]. Only those items applicable to RVI (according to the LRA) were imported into Table B-1 from the LRA. The NUREG-1801 Volume 2 Item, Table 1 Item, and Notes columns are omitted.

Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; Upper Support Plate M-10 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-2 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; Upper Support Column M-10, M-12 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; Upper Support Column Bolts M-10, M-12 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Preload due to Stress Relaxation None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-3 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; Upper Support Column Spider M-10, M-12 Cast Austenitic Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement Loss of Fracture Toughness due to Thermal Embrittlement Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)

Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; Upper Core Plate M-9, M-11 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-4 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; Fuel Alignment Pins M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; Hold-down Spring M-9, M-10 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Preload due to Stress Relaxation None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-5 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; RCCA Guide Tubes M-10 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; RCCA Guide Tube Bolts M-10 Stainless Steel Treated Water (Outside)

Loss of Preload due to Stress Relaxation None Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-6 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; RCCA Guide Tube Support Pins (split pins)

M-10 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; Head and Vessel Alignment Pins M-10 Stainless Steel Treated Water (Outside)

Loss of Material due to Wear ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-7 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper Internals; Head Cooling Spray Nozzles M-11 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Upper Internals; Upper Core Plate Alignment Pins M-10 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Material due to Wear ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-8 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Upper internals; Upper Instrumentation Column, Conduit, and Supports M-12 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Core Barrel M-9, M-11, M-14 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-9 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Core Barrel Flange M-9, M-11, M-14 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Core Barrel Outlet Nozzles M-11 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-10 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Thermal Shield M-14 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Baffle and Former Plates M-9, M-11, M-14 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-11 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Baffle/Former Bolts M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Preload due to Stress Relaxation None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Barrel/Former Bolts M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Preload due to Stress Relaxation None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-12 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Lower Core Plate M-9, M-11, M-12, M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Fuel Alignment Pins M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-13 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Lower Support Forging M-9, M-11, M-12, M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Lower Support Plate Columns M-9, M-12, M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-14 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; BMI Columns M-12 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; BMI Column Cruciforms M-12 Cast Austenitic Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement Loss of Fracture Toughness due to Thermal Embrittlement Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)

Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-15 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Lower Support Plate Column Bolts M-9, M-11, M-12 M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Preload due to Stress Relaxation None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Radial Support Keys M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling None Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Material due to Wear ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-16 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Radial Support Key Bolts M-9 Nickel Base Alloys Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Preload due to Stress Relaxation None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Clevis Inserts M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Material due to Wear ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-17 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Clevis Insert Bolts M-9 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Change in Dimensions due to Void Swelling Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Loss of Preload due to Stress Relaxation None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Tie Plate (Upper and Lower)

M-11, M-12, M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-18 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Diffuser Plate (Note 2)

M-11 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Secondary Core Support M-9, M-11, M-12, M-13 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-19 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Lower Internals; Irradiation Specimen Guide M-4 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Lower Internals; Specimen Plugs M-4 Stainless Steel Treated Water (Outside)

Cracking due to IASCC Cracking due to SCC Water Chemistry Loss of Fracture Toughness due to Neutron Irradiation Embrittlement None Change in Dimensions due to Void Swelling None Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10 CFR 54.21(c).

Loss of Material due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-20 WCAP-18710-NP August 2023 Revision 0 Table B-1: Summary of Aging Management Evaluation - Internals Component Commodity Intended Function (Note 1)

Material Environment Aging Effect Requiring Management Aging Management Program Flux Thimble Guide Tubes M-1, M-4 Stainless Steel Treated Water (Inside)

Cracking due to SCC Water Chemistry Cracking due to Thermal Fatigue TLAA, evaluated in accordance with 10CFR 54.21(c).

Loss of Material Due to Crevice Corrosion Loss of Material due to Pitting Corrosion Water Chemistry Air -Indoor (Outside)

None Non Notes:

(1) Mechanical Intended Function Definitions [1]

M-1 Pressure Boundary M-4 Structural Support M-9 Core Support M-10 Control Rod Support M-11 Core Flow Distribution M-12 Incore Instrumentation Support M-13 Secondary Core Support M-14 Reactor Vessel Shielding (2) The Lower Internals; Diffuser Plate is listed in the LRA, however Westinghouse drawings do not show a Diffuser Plate at Harris, and Duke Energy has verified that Harris does not have a Diffuser Plate [43].

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-1 WCAP-18710-NP August 2023 Revision 0 APPENDIX C MRP-227 AUGMENTED INSPECTIONS Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W1.Control Rod Guide Tube Assembly Guide plates (cards)

All plants (See WEC NSAL 1)

Loss of Material (Wear)

None Per the requirements of WCAP-17451-P, including subsequent examinations.

(Note 5)

Examination coverage per the requirements of WCAP-17451-P, Revision 1.

(Note 5)

See Figure A-2.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-2 WCAP-18710-NP August 2023 Revision 0 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W2.Control Rod Guide Tube Assembly Lower flange welds All plants Cracking (SCC, Fatigue)

Aging Management (IE and TE)

W2.1.Remaining CRGT assembly lower flange welds W2.2.BMI column bodies Enhanced visual (EVT-1) examination to determine the presence of crack-like surface flaws in flange welds no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

100% of outer (accessible)

CRGT lower flange weld surfaces and 0.25-inch of the adjacent base metal on the individual periphery CRGT assemblies.

(Note 2)

See Figure A-3.

W3.Core Barrel Assembly Upper flange weld (UFW)

All plants Cracking (SCC)

W3.1.Upper girth weld (UGW)

W3.3.Lower flange weld (LFW)

W3.2.Upper axial welds (UAW)

W3.4.Lower support forging or casting Enhanced visual (EVT-1) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

100% of the accessible weld length of one side of the UFW and 3/4 of adjacent base metal shall be examined.

(Note 6)

See Figure A-4.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-3 WCAP-18710-NP August 2023 Revision 0 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W4.Core Barrel Assembly Lower girth weld (LGW)

All plants Cracking (SCC, IASCC),

Aging Management (IE)

W4.1.Upper core plate W4.4.Lower support column bodies (cast, non-cast)

W4.2.Middle axial welds (MAW)

W4.3.Lower axial welds (LAW)

Enhanced visual (EVT-1) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

100% of the accessible weld length of the outer diameter (OD) of the LGW and 3/4 of adjacent base metal shall be examined.

(Note 6)

See Figure A-4.

W5.Baffle-Former Assembly Baffle-edge bolts All plants with baffle-edge bolts (Note 11)

Cracking (IASCC, Fatigue) that results in:

  • Lost or broken locking devices
  • Failed or missing bolts

(Note 4)

None Visual (VT-3) examination, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Bolts and locking devices on high-fluence seams. 100% of components accessible from core side.

See Figures A-5, A-6, and A-7.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-4 WCAP-18710-NP August 2023 Revision 0 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W6.Baffle-Former Assembly Baffle-former bolts (Note 7)

All plants (See WEC NSAL 1)

Cracking (IASCC, Fatigue)

Aging Management (IE and ISR)

(Note 4)

W6.2.Lower support column bolts W6.1.Barrel-former bolts Baseline volumetric (UT) examination interval is dependent on the plant design (Note 8).

Subsequent examination is dependent on the plant design and the results of the baseline inspection (Note 9).

100% of accessible bolts.

(Note 3)

See Figures A-5, A-6, and A-7.

W7.Baffle-Former Assembly Assembly (Includes: Baffle plates, baffle edge bolts, corner bolts, and indirect effects of void swelling in former plates)

All plants Distortion (Void Swelling), or Cracking (IASCC) that results in:

  • Abnormal interaction with fuel assemblies
  • Gaps between plates
  • Vertical displacement of baffle plates
  • Broken or damaged edge bolts None Visual (VT-3) examination to check for evidence of distortion, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Core side surface:

  • High fluence baffle joints
  • Top and bottom edge of baffle plates
  • Bolts and locking devices See Figure A-6.
      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-5 WCAP-18710-NP August 2023 Revision 0 Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W8.Alignment and Interfacing Components Internals hold-down spring All plants with 304 stainless steel hold-down springs (Note 12)

Distortion (Loss of Load due to Stress Relaxation)

None Direct measurement of spring height within 3 cycles of the beginning of (before or after) the license renewal period. If the first set of measurements is not sufficient to assess remaining life, additional spring height measurements will be required.

Measurements should be taken at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty.

W9.Thermal Shield Assembly Thermal shield flexures All plants with thermal shields (See WEC TB-19-5)

(Note 13)

Cracking (Fatigue) or Loss of Material (Wear) that results in thermal shield flexures excessive wear, fracture, or complete separation None Visual (VT-3) no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval.

100% of accessible surfaces of 100% of thermal shield flexures.

(Notes 10 and 13)

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-6 WCAP-18710-NP August 2023 Revision 0 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 [30], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [47] and PWROG letter OG-18-46 dated 2/20/2018 [48].

Harris 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 inner diameter (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 [58] 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 [58], 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 [58] (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 [57] and MRP 2017-009 [58]) 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 [19]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

  • Clustering is defined per NSAL-16-1, Revision 1 [56] 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.

(11) Harris does not have baffle-edge bolts [43]; therefore, this item is not applicable to Harris.

(12) The Harris hold-down spring is 403 SS [43]; therefore, this component is not applicable to Harris.

(13) Harris does not have a thermal shield [43]; therefore, this item is not applicable to Harris.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-7 WCAP-18710-NP August 2023 Revision 0 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Control Rod Guide Tube Assembly W2.1.Remaining CRGT lower flange welds All plants Cracking (SCC, Fatigue)

Aging Management (IE and TE)

W2.CRGT Lower Flange Welds Enhanced visual (EVT-1) examination to determine the presence of crack-like surface flaws in flange welds.

Subsequent examination on a 10-year interval.

A minimum of 75% of the CRGT assembly lower flange weld surfaces and 0.25-inch of the adjacent base metal for the flange welds not inspected under the primary link.

See Figure A-3.

Bottom Mounted Instrumentation System W2.2.Bottom-mounted instrumentation (BMI) column bodies All plants Cracking (Fatigue) including the detection of completely fractured column bodies Aging Management (IE)

W2.CRGT Lower Flange Welds Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

100% of BMI column bodies for which difficulty is detected during flux thimble insertion/withdrawal.

See Figure A-9 and A-11.

Core Barrel Assembly W3.1.Upper Girth Weld (UGW)

All plants Cracking (SCC)

W3.Upper Core Barrel Flange Weld (UFW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of one side of the UGW and 3/4 of adjacent base metal shall be examined.

(Notes 2 and 5)

See Figure A-4.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-8 WCAP-18710-NP August 2023 Revision 0 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Core Barrel Assembly W3.2.Upper Axial Weld (UAW)

All plants Cracking (SCC)

W3.Upper Core Barrel Flange Weld (UFW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of one side of the UAW and 3/4 of adjacent base metal shall be examined.

(Notes 2 and 5)

See Figure A-4.

Core Barrel Assembly W3.3.Lower Flange Weld (LFW)

All plants Cracking (SCC)

W3.Upper Core Barrel Flange Weld (UFW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of the OD surface of the LFW and 3/4 of adjacent base metal shall be examined.

(Note 5)

See Figure A-4.

Lower Internals Assembly W3.4.Lower support forging or castings All plants (Note 7)

Cracking (SCC)

Aging Management (TE in Casting)

W3.Upper Core Barrel Flange Weld (UFW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

Minimum of 25% of bottom (non-core side) surface.

(Note 3)

See Figure A-8 and A-9.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-9 WCAP-18710-NP August 2023 Revision 0 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Upper Internals Assembly W4.1.Upper core plate All plants Cracking (Fatigue),

Wear, Aging Management (IE)

W4.Lower Girth Weld (LGW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

Minimum of 25% of core side surfaces.

(Note 3)

See Figure A-12.

Core Barrel Assembly W4.2.Middle Axial Welds (MAW) and W4.3.Lower Axial Welds (LAW)

All plants Cracking (SCC, IASCC)

Aging Management (IE)

W4.Lower Girth Weld (LGW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of the OD of the MAW and LAW and 3/4of adjacent base metal shall be examined.

(Notes 5 and 6)

See Figure A-4.

Lower Support Assembly W4.4.Lower support column bodies (both cast and non-cast)

All plants (Note 8)

Cracking (IASCC)

Aging Management (IE)

W4.Lower Girth Weld (LGW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

25% of the total number of column assemblies (both visible and non-visible from above the 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.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-10 WCAP-18710-NP August 2023 Revision 0 Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Core Barrel Assembly W6.1.Barrel-former bolts All plants Cracking (IASCC, Fatigue)

Aging Management (IE, Void Swelling, and ISR)

W6.Baffle-former bolts (also refer to MRP 2018-002)

Volumetric (UT) examination.

Reinspection every 10 years following initial inspection.

100% of accessible barrel-former bolts (Minimum of 75% of the total population). Accessibility may be limited by presence of thermal shield or neutron pads.

See Figure A-7.

Lower Support Assembly W6.2.Lower support column bolts All plants Cracking (IASCC, Fatigue)

Aging Management (IE and ISR)

W6.Baffle-former bolts Volumetric (UT) examination.

Reinspection every 10 years following initial inspection.

100% of accessible lower support column (LSC) bolts (Minimum of 75% of the total population) or as supported by plant-specific justification.

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

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, Revision 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.

(7) Harris has a lower support forging.

(8) The Harris lower support column bodies are non-cast.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-11 WCAP-18710-NP August 2023 Revision 0 Table C-3: MRP-227, Revision 1-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Reference Examination Method Examination Coverage W10. Core Barrel Assembly Core barrel flange All plants Loss of material (Wear)

ASME Code Section XI Visual (VT-3) examination to determine general condition for excessive wear.

All accessible surfaces at specified frequency.

W11. Upper Internals Assembly Upper support ring or skirt All plants (Note 4)

Cracking (SCC, Fatigue)

ASME Code Section XI Visual (VT-3) examination.

All accessible surfaces at specified frequency.

W12a. Lower Internals Assembly Lower core plate XL lower core plate (Note 1)

All plants Cracking (IASCC, Fatigue)

Aging Management (IE)

ASME Code Section XI as supplemented by TB-16-4 (Note 3)

Visual (VT-3) examination of the lower core plates to detect evidence of distortion and/or loss of bolt integrity.

All accessible surfaces at specified frequency.

W12b. Lower Internals Assembly Lower core plate XL lower core plate (Note 1)

All plants Loss of material (Wear)

ASME Code Section XI as supplemented by TB-16-4 (Note 3)

Visual (VT-3) examination.

All accessible surfaces at specified frequency.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-12 WCAP-18710-NP August 2023 Revision 0 Table C-3: MRP-227, Revision 1-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Reference Examination Method Examination Coverage W13. Bottom-Mounted Instrumentation System Flux thimble tubes All plants Loss of material (Wear)

IEB 88-09 Surface (ET) examination.

Eddy current surface examination as defined in plant response to IEB 88-09.

W14. Alignment and Interfacing Components Clevis bearing STELLITE wear surface Clevis insert bolts (Note 2)

All plants (TB-14-5)

Loss of material (Wear)

Cracking (SCC)

ASME Code Section XI as supplemented by TB-14-5 (Note 2)

Visual (VT-3) examination.

All accessible surfaces at specified frequency.

W15. Alignment and Interfacing Components Upper core plate alignment pins All plants Loss of material (Wear)

ASME Code Section XI as supplemented by TB-16-4 (Note 3)

Visual (VT-3) examination.

All accessible surfaces at specified frequency.

Notes:

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

(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 [34] is not applicable to Harris.

(4) Harris has an upper support skirt.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-13 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W1.Control Rod Guide Tube Assembly Guide plates (cards)

All plants Per the requirements of WCAP-17451-P.

The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

None N/A Per WCAP-17451-P.

W2.Control Rod Guide Tube Assembly Lower flange welds All plants Enhanced visual (EVT-1) examination.

The specific relevant condition is a detectable crack-like surface indication.

W2.1.Remaining accessible CRGT lower flange welds W2.2.Bottom-mounted instrumentation (BMI) column bodies Confirmation of surface-breaking indications in two or more CRGT lower flange welds shall require visual (EVT-1) examination of the remaining accessible CRGT lower flange welds and visual (VT-3) examination of BMI column bodies by the completion of the next refueling outage.

For BMI column bodies, the specific relevant condition for the VT-3 examination is completely fractured column bodies.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-14 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W3.Core Barrel Assembly Upper flange weld (UFW)

All plants Periodic enhanced visual (EVT-1) examination.

The specific relevant condition is a detectable crack-like surface indication.

W3.1.Upper girth weld (UGW)

W3.3.Lower flange weld (LFW)

W3.2.Upper axial welds (UAW)

W3.4.Lower support forging/casting

a. The confirmed detection and sizing of a surface-breaking indication with a length greater than 2 inches in the UFW shall require that the inspection be expanded to include the UGW and LFW by the completion of the next refueling outage.
b. The confirmed detection and sizing of a surface breaking indication with a length greater than 2 inches in either the UGW or LFW shall require that the inspection be expanded to include the UAW by the completion of the next refueling outage.
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.

The specific relevant condition for the expansion core barrel welds (UGW, LFW, UAW) and lower support forging or casting examinations is a detectable crack-like surface indication.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-15 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W4.Core Barrel Assembly Lower girth weld (LGW)

All plants Periodic enhanced visual (EVT-1) examination.

The specific relevant condition is a detectable crack-like surface indication.

W4.1.Upper core Plate W4.4.Lower support column bodies (cast and non-cast)

W4.2.Middle axial welds (MAW)

W4.3.Lower axial welds (LAW)

a. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require inspection of the upper core plate (25%

of the core-side surface) within the next 3 refueling outages. If an indication is found in this inspection of the upper core plate, the examination coverage shall be expanded to 100% of the accessible surface of the core-side surface of the upper core plate during the same refueling outage.

b. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require inspection of the lower support column bodies (cast and non-cast) within the next 3 refueling outages.

The confirmed detection of fractured, misaligned, or missing lower support columns shall require examination of 100% of the accessible uninspected lower support column assemblies using a VT-3 examination from above the lower core plate (minimum of 75%

of the total population of lower support column 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.
a. The specific relevant conditions for the inspection of the upper core plate are broken or missing parts of the plate.
b. The specific relevant conditions for the inspection of the lower support column bodies (cast and non-cast) are fractured, misaligned, or missing columns.
c. The specific relevant condition for the expansion MAW and LAW inspections is a detectable crack-like surface indication.
      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-16 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W5.Baffle-Former Assembly Baffle-edge bolts All plants with baffle-edge bolts (Note 3)

Visual (VT-3) examination.

The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

None N/A N/A W6.Baffle-Former Assembly Baffle-former bolts All plants Volumetric (UT) examination.

The examination acceptance criteria for the UT of the baffle-former bolts shall be established as part of the examination technical justification.

W6.2.Lower support column bolts W6.1.Barrel-former bolts (Note

2)

Confirmation that more than 5% of the baffle former bolts actually examined on the four baffle plates at the largest distance from the core (presumed to be the lowest-dose locations) contain unacceptable indications shall require UT examination of the lower support column bolts within the next 3 fuel cycles.

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.

The examination acceptance criteria for the UT of the lower support column bolts and the barrel-former bolts shall be established as part of the examination technical justification.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-17 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W7.Baffle-Former Assembly Assembly (Includes:

Baffle plates, baffle edge bolts, corner bolts, and indirect effects of void swelling in former plates)

All plants Visual (VT-3) examination.

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence baffle plate joints, 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.

None N/A N/A W8.Alignment and Interfacing Components Internals hold-down spring All plants with 304 stainless steel hold-down springs (Note 4)

Direct physical measurement of spring height.

The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

None N/A N/A

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-18 WCAP-18710-NP August 2023 Revision 0 Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W9.Thermal Shield Assembly Thermal shield flexures All plants with thermal shields (Note 5)

Visual (VT-3) examination.

The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

None N/A N/A Notes:

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

(2) If significant baffle-former bolt clustering (as defined in MRP 2017-009 [58])

is discovered, Harris will implement the Needed requirements of MRP 2018-002 [59] within 3 fuel cycles.

(3) Harris does not have baffle-edge bolts; therefore, this item is not applicable to Harris.

(4) The Harris hold-down spring is 403 SS; therefore, this component is not applicable to Harris.

(5) Harris does not have a thermal shield; therefore, this item is not applicable to Harris.

      • This record was final approved on 8/8/2023, 11:01:15 AM. (This statement was added by the PRIME system upon its validation)

WCAP-18710-NP Revision 0 Non-Proprietary Class 3

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Approval Information Author Approval Young Micah D Aug-07-2023 16:15:44 Verifier Approval Turicik Louis W Aug-08-2023 09:02:18 Reviewer Approval Mckinley Joshua K Aug-08-2023 09:42:42 Manager Approval Musser Kaitlyn M Aug-08-2023 11:01:15 Files approved on Aug-08-2023

U.S. Nuclear Regulatory Commission RA-23-0218 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 CQL-REAC-TM-AA-000001, Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris, Revision 0

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

FORM NUMBER - F-GES-CMPP-PR-DD-000002-1 Rev. 1 DOCUMENT COVER SHEET DOCUMENT NO.

REVISION PAGE 1 of 8 CQL-REAC-TM-AA-000001 0

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ALTERNATE DOCUMENT NUMBER: N/A TITLE: Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris ATTACHMENTS:

© 2022 WESTINGHOUSE ELECTRIC COMPANY LLC, ALL RIGHTS RESERVED - WESTINGHOUSE NON-PROPRIETARY CLASS 3 All Class 3 Documents require a Form 36 to be released.

© 2022 WESTINGHOUSE ELECTRIC COMPANY LLC, ALL RIGHTS RESERVED - WESTINGHOUSE PROPRIETARY CLASS 2 Information included in this material is proprietary and confidential and cannot be disclosed or used for any reason beyond the intended purpose without the prior written consent of Westinghouse Electric Company LLC.

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© 2022 WESTINGHOUSE ELECTRIC COMPANY LLC, ALL RIGHTS RESERVED and/or © 2022 WESTINGHOUSE BUSINESS PARTNER, ALL RIGHTS RESERVED WESTINGHOUSE PROPRIETARY CLASS 2 and/or WESTINGHOUSE BUSINESS PARTNER PROPRIETARY (SEE ATTACHED DOCUMENT)

Information included in this material is proprietary and confidential and cannot be disclosed or used for any reason beyond the intended purpose without the prior written consent of Westinghouse Electric Company LLC.

SUPPLIER OR THIRD PARTY PROVIDED INFORMATION - File And Protect Using Policies For Westinghouse Proprietary Class 2 Information Information included in this material is proprietary and confidential and cannot be disclosed or used for any reason beyond the intended purpose without the prior written consent of the Supplier/Third Party.

Signature Responsibility Name SIGNATURE / DATE(If processing electronic approval select option)

Originator Andrew E. Hawk Electronically Approved***

Reviewer Riley I. Benson Electronically Approved***

Responsible Manager Jesse J Klingensmith Electronically Approved***

(Select one)

(Select one)

  • Approval signifies that the document and all required reviews are complete, the appropriate proprietary class has been assigned, electronic file has been provided to PRIME, and the document is released for use.

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.

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 2 of 8

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

© 2022 Westinghouse Electric Company LLC All Rights Reserved To: David C. Kovacic Date: October 5, 2022 From: Radiation Engineering & Analysis (REA)

Phone: (412) 374-5851 Email: hawkae@westinghouse.com Our Ref: CQL-REAC-TM-AA-000001, Revision 0

Subject:

Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris Reference(s): 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.

3002017168.

Attachment(s): 1. Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris provides a summary of the plant-specific assessment of the MRP-227, Revision 1-A (Reference 1) fuel design and fuel management limitations that was performed to support the development of a reactor internals aging management program (AMP) at Shearon Harris. The results of this assessment show that Shearon Harris meets the fuel design and fuel management limitations of Reference 1. Comments provided on Revision 0-A of this assessment are electronically attached in PRIME.

Please contact the undersigned if there are any questions regarding this information.

Author:

(Electronically Approved)*

Reviewer:

(Electronically Approved)*

Andrew E. Hawk Riley I. Benson Radiation Engineering & Analysis Nuclear Operations Approver:

(Electronically Approved)*

Jesse J Klingensmith Radiation Engineering & Analysis

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 Page 3 of 8

© 2022 Westinghouse Electric Company LLC All Rights Reserved Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations for Shearon Harris

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 Page 4 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 to support the development of a reactor internals aging management program (AMP) at Shearon Harris.

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,
2. criteria defined in MRP-227, Section 2.4, and
3. plant-specific regulatory commitments for managing aging in reactor internals.
      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 Page 5 of 8 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 United States.

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 Shearon Harris was performed in accordance with this guidance.

Plant-Specific Assessment of the MRP-227, Revision 1-A Fuel Design and Fuel Management Limitations Shearon Harris 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 its operating lifetime. This conclusion is based on a comparison of the plant-and cycle-specific core geometries and operating characteristics with the applicability guidelines for Westinghouse-designed reactors specified in Appendix B of MRP-227, Revision 1-A. Note that three different boundaries were explored when developing these guidelines:

1. radial boundary (components laterally surrounding the core),
2. upper axial boundary (components above the core), and
3. lower axial boundary (components below the core).

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

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 Page 6 of 8 characteristics during the second 30 years of operation could impact the MRP-227 reinspection 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:

Shearon Harris transitioned to a low-leakage fuel management strategy with the 3rd fuel cycle following 2.44 calendar years (2.0 effective full-power years (EFPY)) of operation. Since that time, the plant has implemented low-leakage core designs in every fuel cycle except Cycle 11.

There are no current 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) (as defined in Reference 1) shall not exceed 68 W/cm3 Comparison:

For the last five operating fuel cycles (Cycles 19-23), the HGR-FOM at key baffle locations has ranged between 53 W/cm3 and 62 W/cm3. This range of HGR-FOM is representative of anticipated future operation.

Limit 2:

For operation going forward, the average power density of the reactor core (as defined in Reference 1) shall be less than 124 W/cm3.

Comparison:

For the last five operating fuel cycles (Cycles 19-23), Shearon Harris has been operating at the following power level (core power density):

Cycles 19-23: 2948 MWt (111.0 W/cm3)

The power level of 2948 MWt is representative of anticipated future operation.

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 Shearon Harris 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 approximately 12.6 inches. However, the nominal distance between the top of the active fuel and the bottom of the UCP and was less than

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 Page 7 of 8 12.2 inches during Cycles 1 and 2, and Cycles 1 and 2 had a combined length that was greater than 2 calendar years.

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 limit. Therefore, a more detailed evaluation was performed for Shearon Harris to demonstrate that the increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 1 and 2 was more than offset by the margin afforded by the lower operating power density over the entire plant lifetime.

More specifically, the detailed evaluation demonstrated that the core power densities for all completed fuel cycles were less than the 124.0 W/cm3 limit by approximately 10% to 15%.

Since neutron exposure rates are directly proportional to reactor power, these lower, plant-specific core power densities result in a 10% to 15% reduction in neutron exposure rates relative to those that would occur during operation at 124.0 W/cm3.

Conversely, the evaluation also demonstrated that the smaller fuel-assembly-to-UCP gaps experienced during Cycles 1 and 2 were estimated to result in a 2% to 3% increase in neutron exposure rates relative to those that would occur during operation at the 12.2-inch gap limit.

For all other fuel cycles, the evaluations demonstrated that the fuel-assembly-to-UCP gap limit was met with associated margins in the resulting neutron exposure rates ranging from approximately 0.1% to 22%.

When the entire operating lifetime of Shearon Harris is considered, the significantly larger margins associated with the plant-specific core power densities more than offset the increase in neutron exposure rates resulting from the smaller fuel-assembly-to-UCP gaps experienced during Cycles 1 and 2. Therefore, it can be concluded that the plant-specific neutron exposure rates for the 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 limits.

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 Shearon Harris, the rated core power level, including power uprates, has increased from 2775 MWt to 2948 MWt. This variation of rated power level corresponds to a power density range of approximately 104.51 W/cm3 to 111.0 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.

Section 1.0 of this attachment provides the specific criteria in Section 2.4 of MRP-227, Revision 1-A that pertain to fuel design and fuel management. Sections 2.1 and 2.2 of this attachment demonstrate compliance with these criteria.

      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 CQL-REAC-TM-AA-000001, Revision 0 October 5, 2022 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. 3002017168.
      • This record was final approved on 10/7/2022, 11:21:55 AM. (This statement was added by the PRIME system upon its validation)

CQL-REAC-TM-AA-000001 Revision 0 Non-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 Hawk Andrew E Oct-05-2022 09:42:54 Reviewer Approval Benson Riley Oct-05-2022 14:43:43 Manager Approval Klingensmith Jesse J Oct-07-2022 11:21:55 Files approved on Oct-07-2022

U.S. Nuclear Regulatory Commission RA-23-0218 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Regulatory Commitment

U.S. Nuclear Regulatory Commission Page 1 of 1 RA-23-0218 Regulatory Commitment The following table identifies an action committed to in this submittal. Statements in this submittal with the exception of those in the table below are provided for information purposes and are not considered commitments.

Commitment Expected Completion Date Duke Energy commits to update the Shearon Harris Nuclear Power Plant, Unit 1 (HNP),

Updated Final Safety Analysis Report (UFSAR) with a description of the Aging Management Program (AMP) and Inspection Plan for the Reactor Vessel Internals (RVI).

This commitment will be completed in the next periodic update of the UFSAR in accordance with 10 CFR 50.71(e) after NRC approval of the AMP and Inspection Plan for the HNP RVI.

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