ML25149A217
| ML25149A217 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/05/2025 |
| From: | Ilka Berrios Plant Licensing Branch III |
| To: | Jonathan Brown Vistra Operations Company |
| Haeg, LE | |
| References | |
| EPID L-2024-LLR-0037 | |
| Download: ML25149A217 (1) | |
Text
June 5, 2025 Mr. Terry J. Brown Vistra Operations Company LLC c/o Davis-Besse Nuclear Power Station Mail Stop P-DB-3080 5501 N. State Route 2 Oak Harbor, OH 43449-9760
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 - PROPOSED RELIEF REQUEST RR-A1 RELATED TO VESSEL HEAD AND SHELL WELDS AND NOZZLE-TO-VESSEL EXAMINATION COVERAGE REQUIREMENTS (EPID L-2024-LLR-0037)
Dear Mr. Brown:
By letter dated June 5, 2024, as supplemented by the letter dated November 20, 2024, Vistra Operations Company LLC (the licensee) submitted Relief Request (RR) RR-A1 which requests relief from the vessel head and shell welds and nozzle-to-vessel examination coverage requirements specified in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (the ASME Code),Section XI. The request covers inspections conducted during the fourth ten-year Inservice inspection (ISI) interval at the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),
the licensee requested relief and approval to use alternative requirements (if necessary) for ISI items on the basis that the ASME Code requirement is impractical for several components due to access limitations caused by design at the Davis-Besse.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that ASME Code examination coverage requirements are impractical for the subject reactor pressure vessel shell welds and pressurizer shell-to-nozzle welds listed in RR-A1. The NRC staff concludes that based on the volumetric examination coverage obtained, evidence of significant service-induced degradation would have been detected by the examinations that were performed. Furthermore, the NRC staff concludes that the examinations performed provide reasonable assurance of structural integrity of the subject components.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(5). The NRC staff further determines that granting RR-A1 is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Therefore, in accordance with 10 CFR 50.55a(g)(6)(i), the NRC staff grants relief for the subject examinations of the components contained in RR-A1 for the fourth 10-year ISI interval at Davis-Besse, which began on September 21, 2012, and ended on June 7, 2023.
If you have any questions, please contact the Project Manager, Robert Kuntz, at 301-415-3733 or via e-mail at robert.kuntz@nrc.gov.
Sincerely, Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346
Enclosure:
Safety Evaluation cc: Listserv ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.06.05 08:08:16 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF RR-A1 FOR THE FOURTH 10-YEAR INSPECTION INTERVAL VISTRA OPERATIONS COMPANY LLC DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346
1.0 INTRODUCTION
By letter dated June 5, 2024 (Agencywide Documents Access and Management System Accession No. ML24158A004), as supplemented by letter dated November 20, 2024 (ML24326A082), Vistra Operations Company LLC (Vistra, the licensee) submitted relief request (RR) RR-A1 to the U.S. Nuclear Regulatory Commission (NRC, the Commission). RR-A1 requests relief from the vessel head and shell welds and nozzle-to-vessel examination coverage requirements specified in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (the ASME Code),Section XI. RR-A1 covers inspections conducted during the fourth ten-year inservice inspection (ISI) interval at Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),
the licensee requested relief and approval to use alternative requirements (if necessary) for inservice inspection items on the basis that the code requirement is impractical for several components due to access limitations caused by design at Davis-Besse.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the conditions listed therein 10 CFR 50.55a(g)(5)(iii), states, in part that licensees may determine that conformance with certain code requirements is impractical, and shall notify the Commission and submit information in support of the determination.
Regulations in 10 CFR 50.55a(g)(6)(i) state that the Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The licensee has requested relief from the ASME Code,Section XI requirements pursuant to 10 CFR 50.55a(g)(5)(iii). Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the Commission to grant the relief requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Licensees Relief Request 3.1.1 Components Covered under RR-A1 Relief Request RR-A1 covers two reactor vessel welds and four pressurizer nozzle-to-vessel welds.
Weld RC-RPV-WR-34 is a Class 1 reactor vessel lower shell-to-bottom head circumferential weld. It is an Examination Category B-A, Item B1.11 weld. The licensee was only able to obtain 50 percent volumetric coverage. No recordable indications were detected.
Weld RC-RPV-WR-35 is a Class 1 reactor vessel bottom head circumferential weld. It is an Examination Category B-A, Item B1.21 weld. The licensee was only able to obtain 84 percent volumetric coverage. Two recordable indications were discovered, both were found acceptable as subsurface flaws characteristic of slag inclusions from the welding process.
Weld RC-PZR-WP-15 is a Class 1 pressurizer surge nozzle-to-lower head weld. It is an Examination Category B-D, Item B3.110 weld. The licensee was only able to obtain 72.3 percent volumetric coverage. No recordable indications were detected.
Weld RC-PZR-WP-33-W/X is a Class 1 pressurizer W/X axis relief nozzle-to-upper head weld. It is an Examination Category B-D, Item B3.110 weld. The licensee was only able to obtain 52.48 percent volumetric coverage. No recordable indications were detected.
Weld RC-PZR-WP-33-Y/Z is a Class 1 pressurizer Y/Z axis relief nozzle-to-upper head weld. It is an Examination Category B-D, Item B3.110 weld. The licensee was only able to obtain 74.1 percent volumetric coverage. Two recordable indications were detected and found acceptable.
These were the same indications detected during the refueling outage in 2012 and no changes were detected.
Weld RC-PZR-WP-34 is a Class 1 pressurizer spray nozzle-to-upper head weld. It is an Examination Category B-D, Item B3.110 weld. The licensee was only able to obtain 67.60 percent volumetric coverage. No recordable indications were detected.
3.1.2 ASME Code Requirements The code of record for the fourth Ten-Year ISI Interval, which concluded on June 7, 2023, at Davis-Besse is the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with the 2008 Addenda.
Table IWB-2500-1, Examination Category B-A, Item B1.11 requires a volumetric examination of essentially 100 percent of the shell weld length, as defined by Figure IWB-2500-1. ASME Code Case N-460 states that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent.
Table IWB-2500-1, Examination Category B-A, Item B1.21 for reactor vessel circumferential head welds, requires a volumetric examination of essentially 100 percent of the weld length, as defined by Figure IWB-2500-3. ASME Code Case N-460 states that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent.
Table IWB-2500-1, Examination Category B-D, Item B3.110 for pressurizer nozzle-to-vessel welds, requires a volumetric examination of essentially 100 percent of the area of interest, as defined by Figures IWB-2500-7(a), (b), (c), or (d). ASME Code Case N-460 states that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent.
3.1.3 Impracticality of Compliance and Basis for Relief The licensee stated that examination essentially 100 percent of the weld volume or areas of interest for the subject components, as required by the ASME Code, is impractical due to reactor pressure vessel (RPV) internal obstructions and pressurizer nozzle placement inherent to the Babcock & Wilcox vessel designs.
The licensee further stated that for the reactor vessel lower shell to bottom head circumferential weld RC-RPV-WR-34, ultrasonic interrogation of greater than 90 percent of this volume cannot be obtained due to interference caused by the core guide lugs. The core guide lugs are welded to the reactor vessel shell just above the lower shell-to-bottom head weld and extend approximately 2 inches below the centerline of the weld. These lugs restrict the ultrasonic search unit manipulators ability to move to areas necessary to fully examine the required volume.
According to the licensee, for the reactor vessel bottom head circumferential weld number RC-RPV-WR-35, ultrasonic interrogation of greater than 90 percent of this volume cannot be obtained due to interferences caused by reactor vessel incore instrument nozzles and core guide lugs. These instrument nozzles protrude through the bottom head of the reactor vessel to a height of approximately 1 foot from the inside surface of the bottom head.
For the Pressurizer Welds RC-PZR-WP-15, Weld RC-PZR-WP-33-W/X, Weld RC-PZR-WP-33-Y/Z, Weld RC-PZR-WP-34, ultrasonic interrogation of greater than 90 percent of this volume cannot be obtained due to the geometry of the nozzles in conjunction with the pressurizer shell.
3.2
NRC Staff Evaluation
The ASME Code,Section XI, requires essentially 100 percent volumetric examination of pressure retaining welds in the RPV and vessel-to-nozzle welds in the pressurizer. The design configuration of the reactor vessel incore instrument nozzles and core guide lugs in the RPV and the nozzle placement on the pressurizer prevent complete examinations of the subject welds. In order to effectively increase the examination coverage, the RPV core support, instrumentation, and nozzle placement would require design modifications or replacement.
As discussed in the submittal, the licensee obtained 50 percent coverage for the lower shell-to-bottom head circumferential weld RC-RPV-WR-34 and 84 percent coverage for the vessel bottom head circumferential weld RC-RPV-WR-35. The RPV welds were examined using a wide range of inspections angles, including 45 degree shear and longitudinal waves and 70 degree longitudinal waves which were used to detect possible cracks and determine if embedded indications were surface-connected. The licensee reported small indications found in weld RC-RPV-WR-35, however the staff noted that the submittal mentioned no recordable indication for RC-RPV-WR-34.
In a letter dated February 27, 2013 (ML13059A315), FirstEnergy Nuclear Operating Company (the licensee of Davis-Besse at that time), submitted Relief Request L-13-076, which included a notification of impracticality, RR-A36, to obtain the required essentially 100 percent examination coverage of weld RC-RPV-WR-34. In that submittal a recordable indication was reported for weld RC-RPV-WR-34. The indication was found to be acceptable and in the 2013 submittal was described as likely a slag inclusion which had been detected in the previous ISI interval. The staff noted the absence of this indication in the current submittal and sent a request for additional information (RAI) about the condition of the indication (ML24274A148). In its response dated November 20, 2024 (ML24326A082), Vistra stated that the previous 10-year examination results were reviewed prior to the most recent examination and the location of the flaw was examined during the most recent ultrasonic examination of RC-RPV-WR-34. The licensee stated that an indication could not be conclusively detected and recorded in the weld, which the licensee stated is possible due to different technology, recording methodology, and procedure requirements between the scan performed in 2011 and in the current submittal. The NRC staff finds this acceptable because the relevant portion of the weld was examined and the indication had been previously described as a slag inclusion from fabrication, rather than any indication that would result from operation and was previously found acceptable.
In the RAI Responses dated November 20, 2024, the licensee clarified the figures describing the examination coverage for welds WC-RPV-WR-34 and WC-RPV-WR-35. The explanations included descriptions of transducer placement, obstacles to examination, and the examination angle and beam direction. These explanations helped the NRC staff to verify the stated examination coverage and impracticality to achieve full coverage for the two RPV shell welds.
Based on the coverage obtained and the operational experience showing no previous cracking at these and similar circumferential vessel welds in the U.S. reactor fleet, the examinations conducted on the two vessel welds covered by RR-A1 provide reasonable assurance of structural integrity and leak tightness.
The four ASME Class 1, Category B-D, item B3.110 nozzle-to-shell welds for the pressurizer received at least 50 percent coverage in examination. While the RC-PZR-WP-33-Y/Z weld had two recordable indications found during examination, both were found acceptable and neither had grown since examination in the previous ISI interval. Given the examination coverage on all of the subject pressurizer welds, significant patterns of degradation would have been detected if present and reasonable assurance of structural integrity has been provided.
4.0 CONCLUSION
Based on the above, the NRC staff concludes that ASME Code examination coverage requirements are impractical for the subject RPV shell welds and pressurizer shell-to-nozzle welds listed in RR-A1. The NRC staff concludes that based on the volumetric examination coverage obtained, evidence of significant service-induced degradation would have been detected by the examinations that were performed. Furthermore, the staff concludes that the examinations performed provide reasonable assurance of structural integrity of the subject components. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(5). The NRC staff further determines that granting RR-A1 is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, in accordance with 10 CFR 50.55a(g)(6)(i), the NRC staff grants relief for the subject examinations of the components contained in RR-A1 for the fourth 10-year ISI interval at the Davis-Besse, which began on September 21, 2012, and ended on June 7, 2023.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors:
C. Parker, NRR J. Tsao, NRR Date: June 5, 2025
- via eConcurrence NRR-028 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NVIB/BC NAME RKuntz SLent ABuford DATE 5/28/2025 6/2/2025 5/29/2024 OFFICE NRR/DORL/LPL3/BC (A)
NAME IBerrios DATE 6/5/2025