ML25329A322
| ML25329A322 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/25/2025 |
| From: | Sierra C NRC/NRR/DORL/LPL2-2 |
| To: | Duke Energy Progress |
| Purnell, B | |
| References | |
| EPID L-2025-LLR-0092 RA-25-0264 | |
| Download: ML25329A322 (0) | |
Text
VERBAL APPROVAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR A PROPOSED ALTERNATIVE FOR THE ACCEPTANCE OF A THROUGH-WALL FLAW IN A SERVICE WATER SYSTEM VALVE BODY BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DUKE ENERGY PROGRESS, LLC DOCKET NUMBER 50-325 Technical Evaluation read by Matthew Mitchell, Chief of the Piping and Head Penetration Branch, Division of New and Renewed Licenses, Office of Nuclear Reactor Regulation By letter dated November 17, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession ML25321A755) Duke Energy Progress, LLC (Duke Energy, the licensee) submitted request RA-25-0264 to propose an alternative method to calculate the stresses that are used in evaluating a through-wall pinhole flaw identified in the body of Service Water system control valve 1-E11-PDV-F068A, in lieu of repair or replacement in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),
Section XI, IWA-4000. Specifically, the licensee proposed to use the 2007 Edition with 2008 Addenda of the ASME Section III, NB-3200 Design by Analysis approach to demonstrate structural integrity of the subject valve until repair or replacement of the valve can be accomplished during the spring 2028 refueling outage which is scheduled to start on March 4, 2028.
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee proposed the alternative on the basis that compliance with the ASME Code,Section XI repair or replacement requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The alternative is requested to meet the requirement from ASME Code,Section XI, 2007 edition with the 2008 addenda that flaws in ASME Code components which do not meet the standards are to be subjected to a supplemental examination or a repair/replacement activity. In regard to flaw analysis, IWD-3500 states that the requirements of IWC-3500 may be used. However, the Code does not include analytical evaluation criteria for acceptance of through-wall flaws in the pressure-retaining base material of non-ferrous alloy pipe or fittings. ASME Code Case N-869, "Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 or 3 Piping Section XI, Division 1," has been conditionally approved by the U.S. Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 21. This code case provides analytical evaluation rules for temporary acceptance of flaws in piping but does not include evaluation criteria applicable to valve bodies.
As the basis for hardship, the licensee indicated that:
- 1) weld repair of the valve is unfeasible based on prior experience with valves of the same material,
- 2) replacement of the valve cannot be accomplished safely with the unit online,
- 3) replacement of the valve with the unit online would cause the Secondary Containment to be inoperable, and the replacement could likely not be completed within the timeframe required to restore operability of the Secondary Containment, and
- 4) completion of the valve body replacement would require isolation of the Conventional Service Water System discharge header, which would require shutdown of the unit.
The leaking valve is a 16 x 20 angled control valve in the service water system. The valve is fabricated from an aluminum bronze casting conforming to ASME material specification SB-148, Type CD954. The flaw is located on the 16 side of the valve and is currently leaking at a rate of two drops every 30 minutes. The licensee performed an ultrasonic (UT) thickness examination around the area of the leak, which showed no wall thinning, but UT thickness readings could not be obtained in the immediate vicinity of the leak due to the presence of an external gusset. The licensee also performed an internal borescope visual examination that showed no evidence of wall thinning in the area of the leak.
In lieu of performing a weld repair, the licensee performed a stress analysis in accordance with the Design by Analysis approach of Section NB-3200 of ASME Code,Section III, 2007 edition with 2008 addenda. The licensee applied the limit load design criteria of subparagraph NB-3228.1 to evaluate the flaw. The licensee assumed a conservative flaw size and geometry, consisting of a 6.2 axial by 5 circumferential hole, which is larger than the area in which UT thickness readings could not be obtained. The licensee used a 3-D finite element analysis to model the stresses in the valve body, including the assumed flaw. The analysis considered all applicable loadings consisting of Normal/Upset (Service Level A/B) and Faulted (Service Level D) loadings. The analysis showed that the limit load that can be sustained by the valve body including the flaw exceeded the criterion of 1.5 times the allowable stress.
In addition, the licensee indicated that it would perform compensatory actions in accordance with Code Case N-869, including:
- 1) periodic examination at no more than 30-day intervals to determine if the flaw is growing and to establish the time at which the detected flaw will reach the allowable size;
- 2) daily leakage monitoring to confirm the analysis conditions used in the evaluation remain valid (this monitoring will also be performed while the RHR heat exchangers are in service at normal operating temperature and pressure for quarterly testing);
- 3) if engineering evaluation of the examinations reveals flaw growth to be unacceptable, a repair/replacement activity in accordance with IWA-4000 shall be performed; and
- 4) repair replacement activities shall be performed no later than the time when the predicted flaw size from either periodic inspection or by flaw growth analysis exceeds the acceptance criteria, or during the Spring 2028 scheduled refueling outage, whichever occurs first.
The staff notes that per Code Case N-869, Section 5, the periodic examination would consist of a volumetric examination or physical measurement to assess degradation.
The licensee acknowledged that the leak rate from the pinhole flaw is currently low, therefore it does not challenge the structural integrity of the valve, does not affect the required safety function of the valve, or create issues associated with flooding. Leakage from the pinhole is monitored daily to ensure the analysis conditions used in the evaluation remain valid. If the leakage from the pinhole exceeds the 20 gallons per minute (gpm) leak rate limit, the licensee will shut down the unit and replace the degraded valve.
Based on the licensees submittal, the NRC staff understands that:
- 1) the licensee performed a structural analysis demonstrating reasonable assurance of the structural integrity of the valve body until the valve body can be replaced; and
- 2) the licensee will perform compensatory actions in accordance with Code Case N-869 including periodic examination and monitoring.
Therefore, based on the above, the NRC staff finds that (1) there is reasonable assurance that the licensees proposed alternative has a minimal impact on safety and (2) the licensees hardship justification is acceptable.
Based on the above, the NRC staff finds that the licensee has adequately addressed all regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of the analysis method of Section NB-3200 of the 2007 Edition with the 2008 Addenda of the ASME Code,Section III, until the Spring 2028 refueling outage currently scheduled to start on March 4, 2028.
Authorization read by David Wrona, Chief of the Plant Licensing Branch II-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch II-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation, I concur with the Piping and Head Penetration Branchs determinations.
The proposed request provides reasonable assurance of structural integrity of all applicable pressure-retaining piping and components and complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, as of November 25, 2025, the NRC staff authorizes the use of the Design by Analysis approach of Section NB-3200 of ASME Section III 2007 edition with 2008 addenda at Brunswick Steam Electric Plant, Unit No. 1, for the evaluation of the flaw in Service Water system control valve 1-E11-PDV-F068A, until the Spring 2028 refueling outage, currently scheduled to start on March 4, 2028.
All other ASME Code,Section XI requirements for which an alternative was not specifically requested and approved remain applicable including third-party review by the Authorized Nuclear Inservice Inspector.
This verbal authorization does not preclude the NRC staff from asking additional clarifying questions regarding the licensees request while preparing the subsequent written safety evaluation.