ML24116A259
| ML24116A259 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/14/2024 |
| From: | Michael Mahoney Plant Licensing Branch II |
| To: | Haaf T Duke Energy Progress |
| Mahoney M | |
| References | |
| EPID L-2023-LLL-0016 | |
| Download: ML24116A259 (10) | |
Text
May 14, 2024 Thomas P. Haaf Site Vice President Shearon Harris Nuclear Power Plant Mail Code NHP01 5413 Shearon Harris Road New Hill, NC 27562-9300
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - STAFF EVALUATION RELATED TO AGING MANAGEMENT PLAN AND INSPECTION PLAN FOR REACTOR VESSEL INTERNALS (EPID L-2023-LLL-0016)
Dear Thomas Haaf:
By letter RA-23-0218, dated September 21, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23264A034), and supplemented by letter RA-24-0033, dated February 24, 2024 (ML24055A001), Duke Energy Progress, LLC (Duke Energy, the licensee) submitted its Aging Management Program (AMP) and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), reactor vessel internals (RVIs).
The Harris RVI AMP and Inspection Plan is based on Materials Reliability Program (MRP)-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline (ML19339G350), which provides a strategy for managing age-related degradation for RVIs through the period of extended operation, which begins on October 25, 2026.
The NRC staff concludes that the licensee has demonstrated that the Harris RVI AMP and Inspection Plan, as supplemented by letter dated February 24, 2024, will adequately manage the aging effects and provide reasonable assurance of structural integrity of the RVIs through the period of extended operation.
If you have any questions, please contact me at (301) 415-3867 or Michael.Mahoney@nrc.gov.
Sincerely,
/RA/
Michael Mahoney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400
Enclosure:
Staff Evaluation cc: Listserv
Enclosure STAFF EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AGING MANAGEMENT PROGRAM AND INSPECTION PLAN OF REACTOR VESSEL INTERNALS PER MRP-227, REVISION 1-A DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400
1.0 INTRODUCTION
By letter RA-23-0218, dated September 21, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23264A034), and supplemented by letter RA-24-0033, dated February 24, 2024 (ML24055A001), Duke Energy Progress, LLC (Duke Energy, the licensee) submitted its Aging Management Program (AMP) and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), reactor vessel internals (RVIs).
The Harris RVI AMP and Inspection Plan is based on Materials Reliability Program (MRP)-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline (ML19339G350), which provides a strategy for managing age-related degradation for RVIs through the period of extended operation, which begins on October 25, 2026, for Harris.
The licensee submitted its Harris RVI AMP and Inspection Plan in accordance with its Updated Final Safety Analysis Report Regulatory (UFSAR) Section 18.1, Aging Management Programs and Activities, which states, 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 [Harris Nuclear Plant] 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.
2.0 REGULATORY EVALUATION
The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for renewal of operating licenses for nuclear power plants, address the requirements for plant license renewal process. The regulation at 10 CFR 54.21, Contents of application - technical information, requires that each application for license renewal contain an integrated plant assessment and an evaluation of time-limited aging analyses. The plant-specific integrated assessment shall identify and list those structures and components subject to an aging management review and demonstrate that the effects of aging will be adequately managed so that their intended functions will be maintained consistent with the current licensing basis during the period of extended operation as required by 10 CFR 54.29(a).
Structures and components subject to an AMP shall encompass those structures and components that are referred to as passive and long-lived. Passive structures and components perform an intended function, as described in 10 CFR 54.4, without moving parts or without a change in configuration or properties. Long-lived structures and components are not subject to replacement based on a qualified life or specified time period. The scope of components considered for inspection under MRP-227, Revision 1-A, includes core support structures, typically denoted as Examination Category B-N-3 by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4.
The NUREG-1800, Revision 2, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants (ML103490036), includes guidance that NRC staff should follow to confirm that aging degradation effects will be managed for RVI components during the period of extended operation. Some owners of pressurized-water reactors (PWR) units were granted renewed licenses contingent on a commitment to: (1) participate in the industry programs for investigating and managing aging effects on RVI components; (2) evaluate and implement the results of the industry programs as applicable to the RVI components; 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 RVI components to the NRC for review.
3.0 TECHNICAL EVALUATION
3.1 NRC Staff Assessment Approach The NRC staff assessment of the Harris RVI AMP and Inspection Plan focused on determining whether the licensee adequately incorporated the guidelines specified in MRP-227, Revision 1-A, addressed the action item in the NRC safety evaluation (SE) of the topical report, and that the attributes of the Harris RVI AMP are consistent with NUREG-1801 (GALL Report), Section XM.16A, PWR Vessel Internals, as updated via SLR-ISG-2021-01, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors Interim Staff Guidance (ML20217L203).
3.2 Harris RVI AMP and Inspection Plan per MRP-227, Rev 1-A 3.2.1 Applicability of MRP-227, Revision 1-A By letter dated December 3, 2019 (ML19339G350), the Electric Power Research Institute (EPRI) submitted the NRC-approved version, MRP-227, Revision 1-A, which also contains the NRCs SE. By letter dated February 19, 2020 (ML20006D152), the NRC identified items in MRP-227, Revision 1-A, that required clarification. By letter dated May 4, 2020 (ML20127H664),
EPRI clarified these items related to MRP-227, Revision 1-A. By email dated June 9, 2020 (ML20141L313), the NRC accepted and verified that clarification in the EPRI letter represents an addendum or errata of MRP-227, Revision 1-A, and no further changes were needed to be made to the MRP-227, Revision 1-A.
Section 6, MRP-227 Safety Evaluation Conditions and Action Items, of the Harris, RVI AMP and Inspection Plan discusses how MRP-227, Revision 1-A, is applicable to Harris, in terms of the reactor design, RVI material specifications, stresses, and operating conditions. To demonstrate the applicability, Section 6 of the Harris RVI AMP and Inspection Plan discusses how it complies with the general assumptions in Section 2.4, Guideline Applicability, of MRP-227, Revision 1-A. Enclosure 2 to the letter dated September 21, 2023, discusses how Harris complies with the fuel management and design assumption as specified in MRP-227, Revision1-A.
The NRC staff reviewed the Harris RVI AMP and Inspection Plan and Enclosure 2 to the September 21, 2023, letter, and noted the following:
Harris has operated for less than 30 years with a high-leakage core loading pattern.
Harris operates with fuel design and management that meets MRP-227, Revision 1-A, Appendix B, when the entire operating time of the plant since startup was taken into consideration.
Harris currently operates and has operated under base load conditions for the majority of the operating period of the plant.
Plant modifications impacting the Harris RVI made over the operating period of the plant are those specifically directed by Westinghouse, the original equipment manufacturer (OEM).
The licensee confirmed that:
o No additional items that require aging management were identified by comparison with applicable Harris component drawings.
o The materials identified for Harris are consistent with those materials identified in Table 4-4 of MRP-191, Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse [WEC] and Combustion Engineering [CE] PWR Design.
o The Harris RVI internals are the same as, or equivalent to, the typical Westinghouse PWR RVI internals regarding design and fabrication o The Harris RVI materials operated at temperatures within the original design basis parameters o The Harris RVI stress values are consistent with the assumptions in MRP-227, Revision 1-A, since the plant design was maintained by Westinghouse, the OEM, over the operating period of the plant.
Based on its review of the information provided by the licensee, the NRC staff determined that the licensee has appropriately addressed the fuel management and design of the reactor at Harris, such that it complies with the fuel management and design assumption specified in MRP-227, Revision 1-A, and, therefore, are acceptable.
3.2.2 NRC Condition and Licensee Action Item on MRP-227, Revision 1-A The NRC staff accepted MRP-227, Revision 1-A to the extent delineated within the NRC SE, but requested that utilities which reference MRP-227, Revision 1-A address the items described in the letter dated February 19, 2020 (ML20006D152). These items were addressed by the industry within MRP 2020-012, which provides additional clarifications related to contents in MRP-227, Revision 1-A. The NRC staff issued a Topical Report (TR) SE in May 2020 (ML20141L315), which concluded that MRP 2020-012 achieved the following objectives regarding the items raised in the letter dated February 19, 2020:
(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.
In its review of MRP-227, Revision 1-A, the NRC requested that owners who use the TR address the five items described within the NRC letter dated February 19, 2020 (ML20006D152). The licensee stated that out of five items, only items 5a, 5b, 5d, and 5e pertain to Harris. The NRC staff noted that 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 licensee sufficiently clarified Item 5 associated with MRP-227, Revision 1-A in the submittal; therefore, the licensee has resolved the issues raised by the NRC in its letter dated February 19, 2020.
Action Item 1 of Section 4.0 of the NRC SE for MRP-227, Revision 1-A, specifies that licensees shall inspect baffle former bolts and submit evaluations to the NRC. The licensee stated that it has not yet performed baffle-former bolt inspections at Harris, which is categorized as Tier 4 within the Westinghouse Nuclear Safety Advisory Letter (NSAL)-16-1, Revision 1, Baffle-Former Bolts (ML16222A513). The licensee stated that as directed in MRP-227, Revision 1-A, Harris will perform the baseline ultrasonic testing inspection of baffle-former bolts no later than 35 effective full-power years (EFPY) at which time it will determine a baffle-former bolt re-examination period based on inspection findings.
Additionally, if the inspection findings do not meet the examination acceptance criteria defined in Section 5, Examination Acceptance Criteria and Expansion Criteria, of MRP-227, Revision 1-A, the findings will be dispositioned by plant-specific evaluation per Section 7.5, Examination Results Requirement, of MRP-227, Revision 1-A. If aggressive baffle-former bolt degradation is discovered, as defined in MRP-2017-009, Transmittal of [Nuclear Energy Institute] NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev. 1, dated March 15, 2017 (ML17087A106), the licensee stated that 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.
Additionally, the licensee explained that 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 currently applicable baffle-former bolt reinspection interval. The NRC staff noted that both these activities are consistent with Action Item 1 of MRP-227, Rev 1-A.
The NRC staff finds that the licensee will inspect baffle-former bolts in accordance with MRP-227, Revision 1-A, and the applicable interim guidance. The NRC staff noted that the proposed action is consistent with the Action Item 1 of the NRCs SE because: (1) if aggressive baffle-former bolt degradation is discovered, the licensee will submit its evaluation within 1 year following the outage in which degradation was discovered, and (2) if the inspection interval will be lengthened or will exceed the maximum inspection interval recommended in MRP-2017-009, the licensee will submit to the NRC for information at least 1 year prior to the end of the current reinspection interval. The NRC staff finds that the licensee has satisfied Licensee Action Item 1 of the NRC SE of MRP-227, Revision 1-A, and, therefore, the Harris RVI AMP and Inspection Plan is acceptable with respect to Action Item 1.
3.2.3 Harris RVI AMP and Inspection Plan Implementation Schedule Section 7 of the Harris RVI AMP and Inspection Plan, Inspection Plan and Implementation Schedule, addresses the licensees use of Table 4-3 of MRP-227 Revision 1-A to develop its plant-specific inspection plan and implementation timeline. Table 7-1 of the Harris AMP and Inspection Plan, Shearon Harris Unit 1 Primary Component Inspection Plan, discusses the applicable RVI components identified in MRP-227 Revision 1-A Primary components, the examination method, the examination frequency, and the projected EFPY to perform the respective inspection. Additionally, the NRC staff noted that Appendix C of the Harris RVI AMP and Inspection Plan, MRP-227 Augmented Inspections, contains four separate tables (i.e.,
Tables C-1, C-2, C-3, and C-4), that detail the MRP-227 specified item, MRP-227 applicability, aging effect managed, expansion link, examination method/frequency, and examination coverage. The NRC staff reviewed Table 7-1 and Appendix C to ensure the details of the Harris RVI AMP and Inspection Plan and implementation schedule is consistent with the MRP-227, Revision 1-A.
By letter dated May 19, 2023, EPRI MRP issued interim guidance MRP 2023-005, MRP-227 NEI 03-08 Needed Interim Guidance for WEC/CE Core Barrel Inspections (ML23290A020),
based on operating experience identified in November and December 2022 with regards to inspection of the core barrel. The NRC staff noted the Harris RVI AMP and Inspection Plan did not address this interim guidance. In its RAI response dated February 24, 2024 (ML24055A001), the licensee stated that the interim guidance in MRP 2023-005 is applicable to Harris and that it will implement the applicable interim guidance at the time of the next planned RVI core barrel removal coinciding with MRP-227 examinations of the RVIs. The licensee also revised Tables 7-1, C-1, C-2, and C-4 from Westinghouse WCAP-18710-NP, Revision 0, Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals, to address the updated interim guidance in MRP-2023-005. The NRC staff finds the licensees response acceptable because the licensee revised its inspection and implementation plans to ensure adequate inspection of the impacted components of the core barrel during the next opportunity in response to recent operating experience.
Table C-1 of the Harris RVI AMP and Inspection Plan contains the primary inspection and monitoring recommendations for Westinghouse-designed internals for Harris. The NRC staff verified that Table C-1, as revised by letter dated February 24, 2024, is consistent with Table 4-3 of MRP-227, Revision 1-A, as modified by MRP 2023-005. The NRC staff noted that the RVI of Harris does not contain baffle-edge bolts (item number W5), 304 stainless steel hold down springs (item number W8), and thermal shield flexures (item number W9), which are identified in Table 4-3 of MRP-227, Revision 1-A. The NRC staff finds this acceptable as the design of Harris does not include these three components; thus, the inspections per MRP-227, Rev 1-A are not applicable.
Table C-2 of the Harris RVI AMP and Inspection Plan contains the expansion inspection and monitoring recommendations for Westinghouse-designed internals for Harris. The NRC staff verified that Table C-2, as revised by letter dated February 24, 2024, is consistent with Table 4-3 of MRP-227, Revision 1-A, as modified by MRP 2023-005.
Table C-3 of the Harris RVI AMP and Inspection Plan contains the existing inspection and aging management programs credited in recommendations for Westinghouse-designed internals. The NRC staff verified that Table C-3 is consistent with Table 4-3 of MRP-227, Revision 1-A as modified by MRP 2023-005.
Table C-4 of the Harris RVI AMP and Inspection Plan contains the acceptance criteria and expansion criteria recommendations for Westinghouse-designed internals. The NRC staff verified that Table C-3, as revised by letter dated February 24, 2024, is consistent with Table 4-3 of MRP-227, Revision 1-A, as modified by MRP 2023-005. The NRC noted that the RVI of Harris does not contain baffle-edge bolts (item number W5), 304 stainless steel hold down springs (item number W8), or thermal shield flexures (item number W9) which are mentioned in Table 4-3 of MRP-227, Revision 1-A. The NRC staff finds this acceptable as the design of Harris does not include these three components; thus, the inspections, and associated acceptance criteria and expansion criteria, per MRP-227, Rev 1-A are not applicable.
Based on its review, the NRC staff determined that Tables 7-1, C-1, C-2, C-3, and C-4 of Appendix C to the Harris RVI AMP and Inspection Plan are acceptable and consistent with the guidelines of MRP-227 Revision 1-A, as modified by MRP 2023-005.
3.2.4 Reactor Internal Aging Management Program Elements Section 5 of the Harris RVI AMP and Inspection Plan, Shearon Harris Internal Aging Management Program Attributes, states that based on Duke Energys license renewal application commitment dated November 2006 (ML063350262), the Harris RVI AMP and Inspection Plan is credited for managing eight aging degradation mechanisms and their associated effects: stress-corrosion cracking, irradiation-assisted stress corrosion cracking, wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep.
The licensee stated that its Harris RVI AMP and Inspection Plan meets MRP-227, Revision 1-A and is consistent with the recommendations in GALL Report. During its review, the NRC staff reviewed each program element and compared them to with the Section XI.M16A of the GALL Report, as modified by NRC subsequent license renewal Interim Staff Guidance, SLR-ISG-2021-01-PWRVI, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors (ML20217L203), to verify consistency. The NRC staff compared the scope of program, preventative actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmation process, administrative controls, and operating experience program elements of the licensees Harris RVI AMP and Inspection Plan to the corresponding GALL elements in the Section XI.M16A of the GALL Report, as modified by SLR-ISG-2021-01-PWRVI. Based on its review, the NRC staff finds that the ten program elements of the Harris RVI AMP and Inspection Plan are consistent with the corresponding program elements in Section XI.M16A of the GALL Report, as modified by SLR-ISG-2021 PWRVI, and that the licensees AMP is adequate to manage the applicable aging effects of the RVIs at Harris.
4.0 Conclusion The NRC staff has determined that (1) the licensee has adequately addressed Action Item 1 in the NRC SE for MRP-227, Revision 1-A, (2) the Harris RVI AMP and Inspection Plan is consistent with the inspection and evaluation guidance of MRP-227, Revision 1-A, as modified by MRP 2023-005, and (3) the Harris RVI AMP and Inspection Plan is consistent with Section XI.M16A of the GALL Report, as modified by SLR-ISG-2021-01-PWRVI; therefore, it is acceptable.
Based on the above, the NRC staff concludes that the licensee has demonstrated that the Harris RVI AMP and Inspection Plan, as supplemented by letter dated February 24, 2024, will adequately manage the aging effects and provide reasonable assurance of structural integrity of the RVIs through the period of extended operation set to begin on October 25, 2026.
The NRC staff noted that: (1) its approval of the Harris RVI AMP and Inspection Plan does not reduce, alter, or otherwise affect ASME Code,Section XI, inservice inspection (ISI) requirements, or any licensing basis requirements related to the ISI of RVI components at Harris, and (2) if the licensee wishes to use a new version of MRP-227, Revision 1-A, in the future, the new version of the topical report must have prior NRC review and approval.
Contributors: O. Yee, NRR E. Palmer, NRR Date: May 14, 2024
ML24116A259 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DNRL/NVIB/BC(A)
NAME MMahoney ABaxter CFairbanks DATE 04/25/2024 04/30/2024 04/19/2024 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DWrona MMahoney DATE 05/13/2024 05/14/2024