ML23095A297
| ML23095A297 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/14/2023 |
| From: | NRC/NRR/VPOB |
| To: | Southern Nuclear Operating Co |
| Shared Package | |
| ML23095A284 | List: |
| References | |
| EPID L-2022-LLR-0078 | |
| Download: ML23095A297 (1) | |
Text
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE INSPECTION REQUIREMENTS FOR STEAM GENERATOR NOZZLE TO REACTOR COOLANT PUMP CASING WELDS (VEGP 3&4-ISI1-ALT-17)
LICENSEE INFORMATION Licensees: Southern Nuclear Operating Company; Georgia Power Company; Oglethorpe Power Corporation; MEAG Power SPVM, LLC; MEAG Power SPVJ, LLC; MEAG Power SPVP, LLC; City of Dalton, Georgia Licensee Address: 7825 River Road, Waynesboro, GA 30830 Plant Name(s) and Unit(s): Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Docket Nos.: 52-025 and 52-026 APPLICATION INFORMATION Submittal Date: 11/4/2022 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML22308A156 Supplement Date(s): NA Supplement ADAMS Accession No.: NA Licensee Proposed Alternative No. or Identifier: VEGP 3&4-ISI1-ALT-17 Applicable Regulation: Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a(z)(1)
Applicable Code Requirements:
10 CFR 50.55a(g)(3) requires that American Society of Mechanical Engineers (ASME) Code Class 1 components meet the preservice examination requirements set forth in either the edition and addenda of ASME Code,Section XI applied to the construction of the component, or in subsequent editions and addenda that have been incorporated by reference in 10 CFR 50.55a.
10 CFR 50.55a(g)(4) requires that inservice examination of components conducted during the first 10-year inspection interval and subsequent 10-year inspection intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 18 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein.
Enclosure
ASME Code,Section XI, Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires a volumetric and surface examination of all NPS 4 (DN 100) or larger nozzle-to-component butt welds each inspection interval.
10 CFR 50.55a(b)(2)(xlii) imposes conditions on these examinations of nozzle-to-component butt welds, including that the ultrasonic test (UT) examinations meet the requirements of ASME Code,Section XI Mandatory Appendix VIII, that the UT examinations cover 100 percent of the weld volume and that a flaw evaluation is performed of the largest hypothetical crack that could exist in the volume not qualified for UT examination.
Applicable Code Edition and Addenda: The code of record for the first ten-year Inservice Inspection (ISI) Interval is the ASME Code,Section XI, 2017 Edition.
Brief Description of the Proposed Alternative:
The licensee is proposing to perform an encoded volumetric examination of only the inner 1/3 of the weld volume (per the 2017 Edition of ASME Section XI) in lieu of examining the full weld volume required by 10 CFR 50.55a(b)(2)(xiii). The volumetric examination will be performed from the inside diameter (ID) surface. The licensee will also perform the required surface examination on the outside diameter (OD) surface. The UT techniques will be qualified in accordance with the Performance Demonstration Initiative (PDI) Program which satisfies the requirements of ASME Code,Section XI, Appendix VIII, Supplement 10. The licensee also performed an analysis of the largest potential flaw which could be expected to be present in the weld volume that will not be examined to demonstrate that the postulated flaw in the outer 2/3 of the weld would not grow to exceed the size of service flaws allowed by ASME Code,Section XI over the licensed lifetime of the plant.
In addition to the ASME Code,Section XI examinations, the licensee proposes to perform an eddy current examination from the ID surface. Although not an ASME mandatory examination, the eddy current examination utilized on the ID surface will be qualified in accordance with ASME Code,Section V, Article 14 (2017 Edition).
The proposed alternative is requested for the First Ten-Year ISI Interval for Vogtle Electric Generating Plant, Units 3 and 4.
REGULATORY EVALUATION Regulatory Basis: 10 CFR 50.55a(z)(1)
Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements set forth in the ASME Code,Section XI.
Paragraph 10 CFR 50.55a(z) states, in part, that alternatives to the requirements of 10 CFR 50.55a(b)-(h) may be used, when authorized by the Director, Office of Nuclear Reactor Regulation, if (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The licensee has submitted this request on the basis that a proposed alternative would provide an acceptable level of quality and safety.
TECHNICAL EVALUATION This proposed alternative updates a similar alternative for the preservice examinations of the eight subject Steam Generator Nozzle to Reactor Coolant Pump Casing (SG-to-RCP) welds to also apply to the inservice inspection to be performed during the First Ten-Year ISI Interval at Vogtle Electric Generating Plant, Units 3 and 4. The welds are dissimilar metal circumferential butt welds joining the low alloy steel with austenitic stainless steel cladding steam generator nozzle to the cast austenitic stainless steel reactor coolant pump casing. The weld and buttering on the steam generator outlet nozzle are composed of Alloy 690 weld metals.
The first difference between the proposed alternative and the conditions in 10 CFR 50.55a(b)(2)(xlii) is that 10 CFR 50.55a(b)(2)(xlii)(B) requires 100 percent of the weld volume to be examined while the 2017 Edition of ASME Code, Section, XI, Figure IWB-2500-8(f) only requires the inner 1/3 of the weld volume to be examined. To justify this reduction in examination volume, the licensee provided a flaw analysis in letter dated November 4, 2022. Specifically, the licensee described the type and size of fabrication flaws found during radiographic testing and UT of the SG-to-RCP welds. The licensee also provided an analysis to show that postulated flaws in the outer 2/3 of the weld would not grow to exceed the size of service flaws allowed by ASME Code,Section XI over the licensed lifetime of the plant. The staff reviewed the type and size of fabrication flaws found in the welds and determined those found by RT were acceptable per ASME Code,Section III and that those found by UT met the acceptance standards of ASME Code,Section XI. The staff also reviewed the licensees flaw tolerance analysis and determined that it was sufficient to support the alternative because (1) it was performed in accordance with ASME Code,Section XI and (2) the type and size of postulated flaws analyzed were sufficient to conservatively bound the fabrication flaws found in the welds. On this basis, the staff determined that the information provided was sufficient to support the alternative request.
The second difference is that the weld configuration in the licensees alternative request is different than the weld configuration used to illustrate the examination requirements in the 2017 Edition of ASME Code,Section XI. To address the differences between the ASME Code,Section XI, Appendix VIII configurations described and the licensees plant-specific weld configurations, an AP1000 steam generator to pump casing weld specimen was designed and fabricated by the Electric Power Research Institute (EPRI) in accordance with the EPRI/PDI Program. This specimen serves as a blind test specimen necessary to qualify the UT procedure and the UT personnel. The UT procedure and personnel qualifications are conducted by PDI under the PDI ASME Code,Section XI, Appendix VIII program. The examinations will be performed from the ID of the SG-to-RCP welds, which will provide for better detection and sizing than from the OD. Additionally, the licensee will use eddy current on the ID, providing an additional examination to detect any service-induced cracking.
Based on the review described above, the U.S. Nuclear Regulatory Commission (NRC) staff finds that the licensee has demonstrated that the proposed alternative provides an acceptable level of quality and safety. This finding is based on the fact that the licensee has described, in detail, the surface and volumetric examinations to be performed to detect inservice degradation.
For the weld volume that cannot be volumetrically examined, the licensee has provided an analysis to demonstrate that defects in the region of interest will not exceed the flaw size allowed by ASME Code,Section XI over the licensed lifetime of the plant. Finally, the weld and buttering for the welds were fabricated using Alloy 690, which has demonstrated very good
resistance to Primary Water Stress Corrosion Cracking. This information provides a reasonable assurance that the structural and leaktight integrity of the SG-to-RCP welds will be maintained.
CONCLUSION The NRC staff has determined that the proposed alternative in the licensees request referenced above would provide an acceptable level of quality and safety.
The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1).
The NRC staff authorizes the use of proposed alternative VEGP 3&4-ISI1-ALT-17 at Vogtle Electric Generating Plant, Units 3 and 4 for the first Ten-Year ISI Interval.
All other ASME Code,Section XI requirements for which an alternative was not specifically requested and approved in this proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor(s): Stephen Cumblidge, John Honcharik