ML24201A046
| ML24201A046 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/26/2024 |
| From: | Jennivine Rankin Plant Licensing Branch IV |
| To: | South Texas |
| Byrd, Thomas | |
| References | |
| EPID L-2024-LLR-0027, RR-ENG-4-07 | |
| Download: ML24201A046 (1) | |
Text
July 26, 2024 SOUTH TEXAS PROJECT, UNIT 2 - AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NO. RR-ENG-4-07 (EPID L-2024-LLR-0027)
LICENSEE INFORMATION Recipients Name and Address:
Ms. Kym A. Harshaw Acting President, Chief Executive Officer, and Chief Nuclear Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 Licensee:
STP Nuclear Operating Company Plant Name and Unit:
South Texas Project (STP), Unit 2 Docket No.:
50-499 APPLICATION INFORMATION Submittal Date: April 6, 2024 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML24097A007 Applicable Inservice Inspection (ISI) Interval and Interval Start/End Dates: STP, Unit 2, fourth 10-year ISI interval Alternative Provision: The licensee requested an alternative under Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), Hardship without a compensating increase in quality and safety.
ISI Requirement: Paragraph 50.55a(g)(6)(ii)(D), Augmented ISI requirements: Reactor vessel head inspections, of 10 CFR 50.55a, Codes and standards, which specifies the use of American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME Code) Case N-729-6, Alternative Examination Requirements for PWR [Pressurized-Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, with conditions.
Applicable Code Edition and Addenda: 2013 Edition of ASME Code,Section XI Brief Description of the Proposed Alternative: The licensee is proposing an alternative from ASME Code Case N-729-6 paragraphs -3142.2 and -3200 for performing volumetric and/or surface exams of the STP, Unit 2 reactor vessel head (RVH) penetration 75. As an alternative, the licensee proposes to perform a bare metal visual examination (VE) of the RVH surface near penetration 75 at the next refueling outage (2RE24, fall 2025) after cleaning and documenting the as left condition. The licensee explained that performance of a volumetric examination during the spring 2024 refueling outage would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2).
The licensee submitted Relief Request No. RR-ENG-4-07 for the proposed alternative during the spring 2024 refueling outage, 2RE23. The U.S. Nuclear Regulatory Commission (NRC) staff authorized the proposed alternative verbally on April 9, 2024, and documented the script under ADAMS Accession No. ML24101A008. The following safety evaluation provides further written documentation of the NRC staffs regulatory decision.
For additional details on the licensees request, please refer to the documents located at the ADAMS accession numbers identified above.
STAFF EVALUATION In performance of the spring 2024 refueling outage VE of the RVH, the licensee found a relevant condition on the head surface near RVH penetration 75. During light cleaning and evaluation, the licensee took actions to identify the source of the leakage. The licensee concluded that a source of the leakage was from the core exit thermocouple nozzle flange on penetration number 75 above the RVH. However, due to some remaining corrosion products in the annulus region, the licensee was not able to preclude possible nozzle leakage as another source of the leakage. Therefore, the licensee considered the indication on the head surface to be a relevant condition of possible nozzle leakage in accordance with subparagraph -3142.1 of ASME Code Case N-729-6. As such, subparagraphs -3141.2 and -3200 of ASME Code Case N-729-6 require the licensee to perform either a repair/replacement activity or supplemental volumetric and surface examinations. Instead of the supplemental examinations, the licensee proposed the following alternative. The licensee would clean the remaining area of the RVH surface and verify the structural integrity of the RVH. During the upcoming cycle of operation, the licensee stated that it would monitor for leakage in a manner which will continue to ensure the structural integrity of the RVH. Finally, during the next scheduled refueling outage, 2RE24, the licensee would perform an additional VE of the RVH near penetration 75 to ensure no leakage is occurring.
The licensee requested authorization for its proposed alternative under 10 CFR 50.55a(z)(2),
based, in part, on hardship. The NRC staff has reviewed and evaluated the licensees request on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff found the licensees estimation of radiological dose to perform the supplemental volumetric and surface examinations during the spring 2024 refueling outage was consistent with estimates at other facilities. The NRC staff also found that the radiological dose estimate of 9 person-REM for this work is a hardship on the licensee consistent with 10 CFR 50.55a(z)(2).
The NRC staff reviewed the level of quality and safety of the licensees proposed alternative that the volumetric and surface examinations of the subject RVH penetration are not necessary to ensure the structural integrity of the STP, Unit 2, RVH given the licensees proposed alternative.
The licensee provided a supporting basis in noting the penetration nozzle and weld materials resistance to primary water stress corrosion cracking (PWSCC), prior operating history including inspections at STP, Unit 2, defense-in-depth actions including VE exams after cleaning and during the fall 2024 refueling outage, and finally leakage monitoring programs during the subsequent cycle of operation. The NRC staff reviewed each of these factors in evaluating the level of quality and safety in the licensees proposed alternative.
The NRC staff notes that the degradation mechanism of concern is leakage of primary coolant containing boric acid from PWSCC in the RVH penetration and/or associated J-groove weld.
This leakage can cause two issues to challenge the structural integrity of the reactor coolant pressure boundary of the RPV head or nozzles. The first challenge is circumferential cracking, and thereby ejection of a penetration nozzle from the RPV head. This could cause a small break loss-of-coolant accident (LOCA) or control rod misalignment. The second challenge is that the leakage could cause boric acid corrosion of the low alloy steel material that comprises the bulk thickness of the RPV head. Boric acid corrosion rates of low alloy steel could be up to 6 inches/year under very severe conditions as discussed in NUREG/CR-6875, Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials, July 2005 (ML052360563). After sufficient corrosion, a small or medium break LOCA could occur. To prevent significant degradation in RPV heads and penetration nozzles, 10 CFR 50.55a(g)(6)(ii)(D) requires an inspection program for these components, including volumetric and surface examinations when a licensee cannot perform an effective VE of the top of the RVH due to masking of other deposits from above the RVH.
The NRC staff assessed the adequacy of the licensees proposed alternatives and defense-in-depth actions to evaluate the structural integrity of the RVH and penetration 75, specifically. The NRC staff notes that the penetration nozzle and weld material of the RVH at STP, Unit 2 is Alloy 690/52/152. These materials have been extensively tested for PWSCC resistance from crack growth and initiation by NRC contractors. For reference see NUREG/CR-7103, Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys, Volume 1, September 11, 2011 (ML11277A230 and ML11294A228), Volume 2, April 2012 (ML12114A011), Volume 3, July 2016 (ML16190A072),
and Volume 4, April 2019 (ML19099A200); NUREG/CR-7137, Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment - 2009, June 2012 (ML12199A415); and NUREG/CR-7226, Primary Water Stress Corrosion Cracking of High-Chromium, Nickel-Base Welds Near Dissimilar Metal Weld Interfaces, January 2018 (ML18018A562). The NRC staff confirmatory research has verified that the materials are very resistant to PWSCC initiation and crack growth in laboratory testing.
In addition, the NRC staff notes that operating experience has identified no PWSCC during the operational service of these materials in any U.S. PWR, which supports the conclusion that it is unlikely that PWSCC is currently present in the STP Unit 2 RVH penetration 75. Further, if cracking and minor leakage of a RVH penetration was currently occurring, the resistance of Alloy 690/52/152 to crack growth provides additional assurance that any cracking currently present would be unlikely to increase to the point of challenging the structural integrity of the RVH over one additional operating cycle. Finally, due to this slow crack growth rate, the licensees leakage monitoring actions would be able to identify leakage through RVH penetration 75 prior to it presenting a challenge to structural integrity of the RVH. Hence, the NRC staff finds that the licensees proposed alternative provides reasonable assurance of the structural integrity of the RVH for the next operating cycle at STP, Unit 2, without requiring the licensee to perform supplemental volumetric and surface examinations during the spring 2024 refueling outage.
Therefore, the NRC staff finds, given the actions of the licensees proposed alternative under RR-ENG-4-07, there would be limited value in quality and safety in requiring additional supplemental volumetric and surface examinations to verify no indications of cracking in these materials during the current refueling outage. Given the hardship, the NRC staff finds in accordance with 10 CFR 50.55(a)(z) that (1) there is reasonable assurance that the licensees proposed alternative has a minimal impact on quality and safety; and (2) the licensees hardship justification is acceptable.
CONCLUSION As set forth above, the NRC staff has determined that complying with the specified requirements described in the licensees alternative, RR-ENG-4-07, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Further, the licensees proposed alternative provides reasonable assurance of structural integrity of the RVH and penetration 75.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, as of April 9, 2024, the NRC had authorized the use of RR-ENG-4-07 until the next scheduled refueling outage, 2RE24, at STP, Unit 2, scheduled for fall 2025.
All other ASME Code,Section XI requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Jay Collins Date: July 26, 2024 Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation cc: Listserv Jennivine K.
Rankin Digitally signed by Jennivine K. Rankin Date: 2024.07.26 12:46:38 -04'00'
- concurrence via email OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA*
NRR/DNRL/NPHP/BC* NRR/DORL/LPL4/BC*
NAME TByrd PBlechman MMitchell JRankin DATE 7/18/2024 7/18/2024 7/22/2024 7/26/2024