ML23142A273
| ML23142A273 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire  |
| Issue date: | 05/22/2023 |
| From: | Shawn Williams Plant Licensing Branch II |
| To: | Donahue J, Treadway R Duke Energy Corp |
| Jordan, N | |
| References | |
| EPID L-2023-LLR-0003 | |
| Download: ML23142A273 (1) | |
Text
From:
Shawn Williams To:
Treadway, Ryan I Cc:
Vaughan, Jordan L; Sigmon, Chet Austin
Subject:
Duke Fleet - Request for Additional Information RE: Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) (EPID L-2023-LLR-0003)
Date:
Monday, May 22, 2023 1:55:09 PM Attachments:
Duke Fleet RAI - Proposed Alternative for Steam Generator Welds.docx
Dear Mr. Treadway,
By letter dated January 23, 2023 (Agencywide Document Access and Management System Accession Number ML23023A093), Duke Energy Carolinas, LLC (the licensee) submitted a proposed alternative to the inservice inspection requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code regarding the steam generator welds and nozzle inner radius locations at Catawba Nuclear Generation Station Units 1 and 2; McGuire Nuclear Station Units 1 and 2; Oconee Nuclear Station Units 1, 2, and 3; Shearon Harris Nuclear Power Plant Unit 1, and H.B. Robinson Steam Electric Plant Unit 2.
The U.S. Nuclear Regulatory Commission staff has determined that additional information is needed as provided below. A clarification call to ensure mutual understanding was conducted on May 22, 2023.
Please respond by July 7th. Please note that the NRC staffs review is continuing and further requests for information may be developed.
If you have any questions, please contact Shawn Williams at 301-415-1009 or via e-mail at Shawn.Williams@nrc.gov.
Shawn Williams, Senior Project Manager Plant Licensing Branch, II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket Nos.
50-269, 50-270, 50-287, 50-413, 50-414, 50-369, 50-370, 50-400, 50-261
cc: Listserv
REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSPECTION INTERVAL EXTENSION FOR STEAM GENERATOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT 2 DOCKET NOS. 50-413, 50-414, 50-400, 50-369, 50-370, 50-269, 50-270, 50-287, AND 50-261 EPID NO.: L-2023-LLR-0003
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Background===
By letter dated January 23, 2023 (Agencywide Document Access and Management System Accession Number ML23023A093), Duke Energy Carolinas, LLC (the licensee) submitted to the United States Nuclear Regulatory Commission (NRC), a proposed alternative to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) regarding the steam generator (SG) welds and nozzle inner radius locations at Catawba Nuclear Generation Station Units 1 and 2; McGuire Nuclear Station Units 1 and 2; Oconee Nuclear Station Units 1, 2, and 3; Shearon Harris Nuclear Power Plant Unit 1, and H.B. Robinson Steam Electric Plant Unit 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 55a, Paragraph (z)(1) (10 CFR 50.55a(z)(1)), the licensee is proposing to eliminate the required volumetric examinations of the subject SG welds from the current ASME Code,Section XI 10-year requirement, for varying lengths of time, depending on the site. The licensee referred to the results of the probabilistic fracture mechanics (PFM) analyses in the following Electric Power Research Institute (EPRI) non-proprietary reports as the primary basis for the elimination of the ISI examinations:
EPRI Technical Report 3002015906, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 2019 (hereinafter referred to as EPRI report 15906, ADAMS Accession No. ML20225A141).
EPRI Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, 2019 (hereinafter referred to as EPRI report 14590, ADAMS Accession No. ML19347B107).
The NRC staff needs additional information to complete its review and approval of the licensees submittal.
Regulatory Basis The NRC has established requirements in 10 CFR Part 50 to protect the structural integrity of structures and components in nuclear power plants. Among these requirements are the ISI requirements of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a to ensure that adequate structural integrity of SG vessels (including their welds) is maintained through the service life of the vessels. Therefore, the regulatory basis for the following requests for additional information (RAIs) has to do with demonstrating that the proposed alternative ISI requirements would ensure adequate structural integrity of the licensees SG welds, and thereby would provide an acceptable level of quality and safety per 10 CFR 50.55a(z)(1).
RAI-1
Issue The licensee referenced probabilistic and deterministic analyses in the above EPRI reports to estimate potential fatigue crack growth in the subject SG welds and to justify application of these analyses to the proposed examination elimination for the welds and nozzle inner radius locations. The licensee presented plant-specific information to demonstrate that the referenced analyses in the EPRI reports would bound the subject SG welds, including the ISI history of the welds.
Leveraging PFM analyses to define the basis for risk-informing inspection requirements requires knowledge of both the current and future behavior of the material degradation and the associated uncertainties applicable to the subject SG welds. Confidence in the results of these analyses hinges on the assurance that the PFM model adequately represents, and will continue to represent, the degradation behavior in the subject SG welds. The NRC staff has determined that, when considering proposed elimination of examinations, adequate performance monitoring through inspections is needed to ensure that the assumptions of the PFM model remain valid, and that novel or unexpected degradation is detected and dispositioned in a timely fashion.
Further, the NRC staff has communicated concepts that licensees can implement on a fleet-wide basis to develop a performance monitoring plan and bolster the technical basis for alternative requests (see slide packages dated January 30, 2023, and April 27, 2023 at ML23033A667 and ML23114A034, respectively). In Section 5.0 of the submittal, the licensee described the various plant-specific examination scenarios and the proposed elimination of examinations. The licensee stated that the proposed alternative results in a maximum time period of approximately 20 years from the end of the interval in which the Section XI requirements were met in full until the end of the proposed alternative. The licensee did not provide a performance monitoring schema for the subject welds and nozzle inner radius locations.
The licensee discusses the system leakage test as providing further assurance of safety for the proposed alternative. However, the NRC staff notes that the visual examinations performed during system leakage tests may not provide sufficient information to ensure that the PFM model continues to predict the material behavior and that emergent degradation is discovered and dispositioned in a timely fashion. Specifically, visual examinations may not directly detect the presence or extent of degradation; may not provide direct detection of aging effects prior to potential loss of structure or intended function; and do not provide sufficient validating data necessary to confirm the modeling of degradation behavior in the subject welds and nozzle inner radius locations.
Request
- a. Describe the performance monitoring that will be implemented with this proposed alternative to ensure that the PFM model adequately represents, and will continue to represent the degradation behavior in the subject components commensurate with the duration of the requested alternative.
- b. Explain how this performance monitoring will provide, over the extended examination interval, (1) direct evidence of the presence and extent of degradation, (2) validation and confirmation of the continued adequacy of the PFM model; and (3) timely detection of novel or unexpected degradation.
- c. If through this performance monitoring indications are detected that exceed the acceptance standards of ASME Code,Section XI, IWB-3500, confirm that they will be evaluated as required by ASME Code,Section XI (which includes requirements for successive inspections and additional examinations) and describe other actions (if any) specified in the plants corrective action program to ensure that the integrity of the component is adequately maintained.
- d. If through this performance monitoring indications are detected that exceed the acceptance standards of ASME Code,Section XI, IWB-3500, then scope expansion may be appropriate to assess extent of condition. Furthermore, if this performance monitoring plan or industry-wide operating experience indicates that a new or novel degradation mechanism is possible in SG welds or nozzle inner radii, scope expansion may be appropriate to ensure that no such mechanism is occurring in the subject plants. Discuss the detailed scope expansion plans for these scenarios.
RAI-2
Issue Table 6-4 in Attachment 6 of the submittal noted Value not available for the 60-year projected cycles for the Loss of Power transient at H.B. Robinson Steam Electric Plant.
Request Confirm that the 60 cycles analyzed in EPRI report 14590 for the Loss of Power transient reasonably bound any occurrence of the transient that might occur or could have occurred at H.B. Robinson Steam Electric Plant.
RAI-3
Issue An ISI interval at a particular plant site may be extended, within certain limitations, per IWA-2430 or similar provisions of Section XI. As such, interval dates may extend beyond the end of the operating license for the plant. The NRC may not approve a proposed alternative beyond the end of the current license.
Request Confirm that the proposed alternative does not apply beyond the current licenses of the subject plants.