ML25226A157
| ML25226A157 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  (DPR-082) |
| Issue date: | 08/21/2025 |
| From: | Tony Nakanishi Plant Licensing Branch IV |
| To: | Gerfen P Pacific Gas & Electric Co |
| Lee S, 301-415-3158 | |
| References | |
| EPID L-2025-LLR-0055 | |
| Download: ML25226A157 (1) | |
Text
August 21, 2025 Ms. Paula Gerfen Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 - AUTHORIZATION OF REQUEST FOR ALTERNATIVE NDE-U2-RPV-20-YEAR (EPID L-2025-LLR-0055)
Dear Ms. Gerfen:
By letter dated May 22, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25142A381), Pacific Gas and Electric Company (PG&E, the licensee) submitted request NDE-U2-RPV-20-YEAR, which proposed an alternative to the inservice inspection (ISI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-2411 for Diablo Canyon Nuclear Power Plant (Diablo Canyon), Unit 2.
This section requires volumetric examination of essentially 100 percent of reactor pressure retaining welds identified in table IWB-2500-1 once each 10-year interval. Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested the use of a proposed alternative to extend the Diablo Canyon, Unit 2, reactor vessel fourth inspection interval from 10 years to 20 years on the basis that the alternative provides an acceptable level of quality and safety.
Based on the enclosed safety evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff has concluded that extending the fourth ISI interval for Categories B-A and B-D components from 10 to 20 years will not result in any considerable increase in risk. This conclusion relies on the basis that the Diablo Canyon, Unit 2, reactor pressure vessel is bounded by Topical Report WCAP--16168-NP-A, Revision 3, and the request met all of the provisions set forth in WCAP-16168-NP-A, Revision 3, as described in the NRC staffs safety evaluation dated July 26, 2011. Therefore, the proposed alternative will provide an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Pursuant to 10 CFR 50.55a(z)(1),
the staff concludes that the licensees alternative ISI schedule for the specified welds is acceptable for extension consistent with the schedule in the Pressurized Water Reactor Owners Group letter OG-10-238 dated July 12, 2010. Therefore, the examination of the Categories B-A and B-D components for Diablo Canyon, Unit 2, shall be conducted prior to the end of the extended fourth interval.
P. Gerfen All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the Diablo Canyon project manager at 301-415-3168 or via email at Samson.Lee@nrc.gov.
Sincerely, Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-323
Enclosure:
Safety Evaluation cc: Listserv TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.08.21 16:17:18 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE NDE-U2-RPV-20-YEAR PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NCULEAR POWER PLANT, UNIT 2 DOCKET NO. 50-323
1.0 INTRODUCTION
By letter dated May 22, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25142A381), Pacific Gas and Electric Company (PG&E, the licensee) submitted request NDE-U2-RPV-20-YEAR, which proposed an alternative to the inservice inspection (ISI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-2411 for Diablo Canyon Nuclear Power Plant (Diablo Canyon or DCPP), Unit 2. This section requires volumetric examination of essentially 100 percent of reactor pressure retaining welds identified in table IWB-2500-1 once each 10-year interval. Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(z)(1), Acceptable level of quality and safety, the licensee requested the use of a proposed alternative to extend the Diablo Canyon, Unit 2, reactor vessel fourth inspection interval from 10 years to 20 years on the basis that the alternative provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
Inservice inspection of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda, as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements, except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i), Impractical ISI requirements: Granting of Relief.
The regulation in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, states:
Alternatives to the requirements of paragraphs (b) through (h) of [10 CFR 50.55a]
or portions thereof may be used, when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:
(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality or safety. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirements for operating plants, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME Code, that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of 10 CFR 50.55a, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations in 10 CFR 50.55a(g) require that inservice examination of components and system pressure tests conducted during the successive 120-month inspection intervals (following the initial 120-month inspection interval) must comply with the requirements in the latest edition and addenda of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(a), twelve months before the start of the 120-month interval [or the optional ASME Code Cases listed in NRC Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17, August 2014 (ML13339A689),
subject to the conditions listed in 10 CFR 50.55a(b), Use and conditions on the use of standards].
RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988 (ML003740284), describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor pressure vessels (RPVs).
RG 1.174, Revision 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, November 2002 (ML023240437),
describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.
The licensee has requested an alternative to the ASME Code requirements pursuant to 10 CFR 50.55a(z)(1). The Diablo Canyon, Unit 2, fourth 10-year lSI interval is based on the ASME Code,Section XI, 2007 Edition with 2008 Addenda. The applicable Code for the fifth 10-year ISI interval will be the ASME Code,Section XI, 2019 Edition. 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 authorize, the alternative proposed by the licensee.
The end date for the current Diablo Canyon, Unit 2, fourth 10-year interval lSI program is March 12, 2026.
3.0 BACKGROUND
The ISI of Categories B-A and B-D components consists of visual and ultrasonic examinations intended to discover whether new flaws have initiated, whether preexisting flaws have extended, and whether preexisting flaws may have been missed in prior examinations. These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code.
3.1 WCAP-16168-NP-A, Revision 3 By letter dated May 8, 2008 (Package ML081060053), the NRC staff issued a final safety evaluation (SE), which found that Topical Report WCAP-16168-NP, Revision 2 (the WCAP),
Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, is acceptable for referencing in licensing applications for pressurized water reactors (PWRs) designed by Westinghouse Electric Company (Westinghouse), Combustion Engineering, Inc., and Babcock and Wilcox, Inc. (B&W). The Westinghouse Commercial Atomic Power report (WCAP) was developed to support a risk-informed assessment of extensions to the lSI intervals for ASME Code,Section XI, Examination Categories B-A and B-D components, from 10 to 20 years using data from three different PWR plants (referred to as the pilot plants) representing each of the vendors.
The analyses in the WCAP used probabilistic fracture mechanics tools and inputs from the work described in NUREG-1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report, August 2007 (ML072830076) and NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), March 2010 ML15222A848). The PWR Owners Group (PWROG) analyses incorporated the effects of fatigue crack growth and lSI data. Design basis transient data was used as an input for the fatigue crack growth evaluation. The effects of lSI data were modeled consistently with the previously-approved probabilistic fracture mechanics codes contained in WCAP-14572-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection (ML012630327, ML012630349, and ML012630313). These effects were inputs into the evaluations performed with the Fracture Analysis of Vessels - Oak Ridge (FAVOR) computer code. All other inputs were identical to those used in the PTS risk reevaluation underlying 10 CFR 50.61a, Alternative fracture toughness requirements for protection against pressurized thermal shock events.
The PWROG concluded, as a result of these studies, that the ASME Code,Section XI, 10-year lSI interval for Examination Categories B-A and B-D components in PWR RPVs can be safely extended from 10 to 20 years. This conclusion, based on the results from the pilot plant analyses, was considered to apply to any plant designed by the three PWR vendors represented in the pilot plant study, as long as certain critical plant-specific criteria (defined in Appendix A of the WCAP) are bounded by the analysis for the applicable pilot plant.
3.1.1 Summary of NRC Staff Evaluation for WCAP-16168-NP-A, Revision 3 The NRC staff issued a second SE (ML11306A084), superseding the initial SE in the WCAP, which addressed the PWROGs request for clarification of the information needed in applications utilizing the WCAP. In the NRC SE for WCAP-A, the staff concluded that the methodology presented in the WCAP is consistent with the guidance provided in RG 1.174, Revision 1, and is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and conditions specified in the SE. In addition to confirming that the subject plant is bounded by the pilot plants/parameters identified in Appendix A in the WCAP, licensees that submit a request for an alternative based on the WCAP need to submit the following plant-specific information:
- 1. Licensees must demonstrate that the embrittlement of their RPV is within the envelope used in the supporting analyses. Licensees must provide the 95th percentile total through-wall cracking frequency (TWCFTOTAL) and its supporting material properties at the end of the period in which the relief is requested to extend
the ISI from 10 to 20 years. The 95th percentile total TWCF (TWCF95-TOTAL) must be calculated using the methodology in NUREG-1874. The values for RTMAX-X and the shift in the Charpy transition temperature due to irradiation defined at the 30 ft-lb energy level, T30, must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRC-approved methodology. RTMAX-X is the material property which characterizes the reactor vessels resistance to fracture initiating from flaws in plates (RTMAX-PL), forgings (RTMAX-FO), axial welds and circumferential welds (RTMAX-AW/CW).
- 2. Licensees must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in the PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.
- 3. Licensees must report the results of prior ISI of RPV welds and the proposed schedule for the next 20-year ISI interval. The 20-year inspection interval is a maximum interval. In its request for an alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-10-238 dated July 12, 2010 (ML11153A033).
- 4. Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bounds the fatigue crack growth for all of its design basis transients and (b) identify the design bases transients that contribute to significant fatigue crack growth.
- 5. Licensees with RPVs having forgings that are susceptible to underclad cracking and with RTMAX-FO values exceeding 240 degrees Fahrenheit (°F) must submit a plant-specific evaluation to extend the inspection interval for ASME Code,Section XI, Categories B-A and B-D RPV welds from 10 to a maximum of 20 years because the analyses performed in the WCAP are not applicable.
- 6. Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61a.
WCAP-16168-NP-A, Revision 3, which contains this SE for the WCAP, was issued in October 2011 (ML11306A084, referred to as the WCAP-A in the rest of this SE).
3.2 Proposed Alternative 3.2.1 Description of Proposed Alternative The licensee proposes to defer the ASME Code required Categories B-A and B-D weld ISI for Diablo Canyon, Unit 2, for an additional 10 years. The licensee stated that the proposed ISI dates are consistent with the schedule proposed in PWROG Letter OG-10-238, the latest NRC staff reviewed implementation plan for the PWROG plants.
3.2.2 Components for Which Relief is Requested The affected component is the Diablo Canyon, Unit 2, RPV. The following examination categories and item numbers from table IWB-2500-1 of the ASME Code,Section XI, are addressed in this request:
Examination Category Item Number Description B-A B1.10 Shell Welds B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.20 Head Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section 3.2.3 Basis for Proposed Alternative The basis for the proposed alternative is WCAP-A. Plant-specific parameters for Diablo Canyon, Unit 2, are summarized in tables 1, 2 and 3 of the submittal. All of the critical parameters listed in tables 1, 2, and 3 submittal are bounded by the WCAP-A Westinghouse pilot plant.
3.3 Staff Technical Evaluation The NRC staff reviewed the licensees proposal to extend the Diablo Canyon, Unit 2, ISI interval in order to determine whether the licensee met the risk-informed criteria set forth in the WCAP-A for a Westinghouse plant. By demonstrating that Diablo Canyon, Unit 2, is bounded by the Westinghouse pilot plant analysis with respect to the six criteria identified in section 3.1.1 of this SE, the licensee would have a sufficient technical basis for extending the ISI in accordance with the provisions of the WCAP-A. The Diablo Canyon, Unit 2, RPV has a single layer cladding and is bounded by the Westinghouse pilot plant basis.
The licensee proposed an examination date consistent from the latest reviewed implementation plan, in Letter OG-10-238 for the PWROG plants.
Addressing the first criteria in the SE for WCAP-A, table 3, Details of TWCF Calculation for DCPP Unit 2, at 54 Effective Full-Power Years (EFPY), the licensees submittal provided the TWCF of the limiting axial weld, plate, and circumferential weld, as well as the critical parameters needed to perform the calculations. In order to calculate the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, T30, the licensee used the methodology provided in RG 1.99, Revision 2. The licensee reported that the TWCF95-TOTAL for Diablo Canyon, Unit 2, was 2.5610 per year, which is well within the Westinghouse pilot plant requirement of less than 1.76 10 per year. The NRC staff performed an independent analysis which verified the results reported by the licensee; therefore, the NRC staff finds the TWCF95-TOTAL acceptable, therefore meeting the first criterion With regard to the second criterion in the SE for WCAP-A, the frequency and severity of design basis transients, the licensee was required to show that Diablo Canyon, Unit 2, has a number of
heatup/cooldown transients bounded by that of the Westinghouse pilot plant basis (seven heatup/cooldown cycles per year). The projected number of reactor coolant system transient cycles for 60 years of operation is provided in table 4.3-1 of the Diablo Canyon, Units 1 and 2, license renewal application dated November 7, 2023 (ML23311A154). Table 4.3-1 indicates that Diablo Canyon, Unit 2, is projected to have 71 heatups and 73 cooldowns of the reactor coolant system, over 60 years of operation. These values are well below the 7 heatup/cooldown cycle per year threshold, therefore the NRC staff agrees that the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the Westinghouse design basis transients identified in the PWROG fatigue analysis. Therefore, the second criterion is met.
For the third criterion, the licensee stated that three complete 10-year ISIs have been performed on Diablo Canyon, Unit 2. Indications identified during the most recently completed ISI were identified in table 2, Additional Information Pertaining to the RPV Inspection for DCPP Unit 2, of the submittal. One indication was identified in weld material and one indication was identified in plate material in the beltline region. The indications were found to be acceptable in accordance with table IWB-3510-1 of the ASME Code,Section XI. Since both indications were within the inner 1/10th of the reactor vessel thickness or inner 1 inch of the RPV wall thickness, the indications were evaluated per the requirements of the Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a). Assessment of the two indications against the allowable scaled maximum number of occurrences per 14,560 square-inches per the requirements of the Alternate PTS Rule were found to be acceptable. This information satisfies the third criterion.
Diablo Canyon, Unit 2, is a Westinghouse plant so the fourth criterion related to the bounding fatigue crack growth for all design basis transients and identification of design basis transients that contribute to significant fatigue crack growth in B&W plants is not applicable.
The fifth criterion stated in section 3.1.1 of this SE, requires that plants with forgings that are susceptible to underclad cracking and with RTMAX-FO values exceeding 240 °F must submit a plant-specific evaluation to extend the inspection interval because the analyses performed in the WCAP-A are not applicable. The Diablo Canyon, Unit 2, RPV is fabricated with plates and axial and circumferential welds. Since there are no beltline forging materials, the fifth criterion also is not applicable to this plant.
Lastly, the licensee is not currently seeking additional interval extensions, so the sixth and final criterion is not applicable.
In summary, the licensees submittal demonstrated that the RPVs for Diablo Canyon, Unit 2, is bounded by the Westinghouse limitations set forth in the WCAP-A and the associated SE by the NRC staff. Therefore, the NRC staff concludes that the licensee has adequately demonstrated that the Diablo Canyon, Unit 2, RPV meets all of the applicable criterion set forth in the WCAP-A.
4.0 CONCLUSION
The NRC staff has completed its review of the licensees submittal for an alternative ISI extension to allow use of alternate reactor inspection interval requirements for Diablo Canyon, Unit 2. The NRC staff concluded that extending the fourth ISI interval for Categories B-A and B-D components from 10 to 20 years will not result in any considerable increase in risk. This conclusion relies on the basis that the Diablo Canyon, Unit 2, RPV is bounded by the WCAP-A and the request met all of the provisions set forth in the WCAP-A and as described in the NRC
staffs January 26, 2011, SE for the WCAP-A. Therefore, the proposed alternative will provide an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1).
Pursuant to 10 CFR 50.55a(z)(1), the NRC staff concludes that the licensees alternative ISI schedule for the specified welds is acceptable for extension consistent with the schedule in Letter OG-10-238. Therefore, the examination of the Categories B-A and B-D components for Diablo Canyon, Unit 2, shall be conducted prior to the end of the extended fourth interval.
All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Carolyn Fairbanks Date: August 21, 2025
- by eConcurrence OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA*
NRR/DNRL/NVIB/BC*
NAME SLee PBlechman ABuford DATE 8/14/2025 8/18/2025 8/20/2025 OFFICE NRR/DORL/LPL/BC*
NRR/DORL/LPL4/PM*
NAME TNakanishi SLee DATE 8/21/2025 8/21/2025