ML25044A033
| ML25044A033 | |
| Person / Time | |
|---|---|
| Site: | Farley, Vogtle  |
| Issue date: | 04/08/2025 |
| From: | Markley M NRC/NRR/DORL/LPL2-1 |
| To: | Coleman J Southern Nuclear Operating Co |
| Lamb J, NRR/DORL/LPL2-1 | |
| References | |
| EPID L-2024-LLR-0047 | |
| Download: ML25044A033 (1) | |
Text
April 8, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company 3535 Colonnade Parkway, Birmingham, AL 35243
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, AND VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 - REQUEST FOR ALTERNATIVE REQUIREMENTS FOR PROPOSED INSERVICE INSPECTION ALTERNATIVE GEN-ISI-ALT-2024-03 FOR PRESSURIZER WELDS (EPID L-2024-LLR-0047)
Dear Jamie Coleman:
By letter dated July 3, 2024, (Agencywide Documents Access and Management System Accession No. ML24185A245), as supplemented by letter dated November 15, 2024 (ML24320A121), Southern Nuclear Operating Company (SNC, the licensee) submitted requests NL-24-0227 for U.S. Nuclear Regulatory Commission approval of proposed inservice inspection (ISI) alternative GEN-ISI-ALT-2024-03 for Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, and Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, in accordance with Title 10, Code of Federal Regulations (10 CFR), Section 50.55a(z)(1). This proposed alternative would increase the inspection interval of American Society of Mechanical Engineers (ASME)
Section XI IWB-2500(a), Table IWB-2500-1 Examination Categories B-B and B-D for Item numbers B2.11, B2.12, and B3.110 from every ISI interval to every other interval.
Specifically, pursuant to 10 CFR 55.55a(z)(1), the licensee requested that the NRC authorize Alternative Request GEN-ISI-ALT-2024-03 to defer the ISI examinations for certain pressurizer welds from the current ASME Code,Section XI ISI interval requirement to every other interval.
The regulation in 10 CFR 50.55a(z)(1) requires SNC to demonstrate that the proposed alternative provides an acceptable level of quality and safety.
The NRC staff has reviewed the subject request for relief and concludes, as set forth in the enclosed safety evaluation, that the proposed alternative, GEN-ISI-ALT-2024-03 provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes proposed alternative GEN-ISI-ALT-2024-03 through the Sixth10-year ISI interval that ends June 25, 2037, for Farley, Unit 1; and November 30, 2037, for Farley, Unit 2; and through the Fifth 10-year ISI interval that ends May 30, 2037, for Vogtle, Units 1 and 2. All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief from, or an alternative to, was not specifically requested approved in this subject request remain applicable.
If you have questions, please contact the Senior Project Manager, John Lamb, at 301-415-3100 or John.Lamb@nrc.gov.
Sincerely, Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulations Docket Nos.: 50-348, 50-364, 50-424, and 50-425
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.04.08 08:00:48 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ALTERNATIVE REQUEST GEN-ISI-ALT-2024-03 SOUTHERN NUCLEAR OPERATING COMPANY, INC JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 DOCKET NOS. 50-348, 50-364, 50-424, AND 50-425
1.0 INTRODUCTION
By letter dated July 3, 2024 (Agencywide Documents Access and Management System Accession No. ML24185A245), and supplemented by letter dated November 15, 2024 (ML24320A121), Southern Nuclear Operating Company (SNC, the licensee) submitted a request for Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, and Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, to the U.S. Nuclear Regulatory Commission (NRC) for a proposed alternative to defer certain American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV Code),Section XI examinations.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
SNC proposed to increase the ASME Code,Section XI, inservice inspection (ISI) interval for the requested components, thereby deferring the examinations for the respective components to every other ISI interval on the basis that the proposed alternative provides an acceptable level of quality and safety. In its letter dated July 3, 2024, as supplemented by letter dated November 15, 2024, the licensee requested an alternative based on plant-specific applicability of a technical report prepared by the Electric Power Research Institute (EPRI). The NRC staff reviewed the proposed alternative request for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, as a plant-specific alternative. The NRC did not review the EPRI report for generic use, and this approval does not extend beyond the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, plant-specific authorization.
The proposed alternative also contains details regarding the potential plant-specific applicability of the implementation of ASME Code Case N-921 at Farley, Units 1 and 2, and Vogtle, Units 1 and 2, which extends ISI intervals from 10-years to 12-years. This safety evaluation (SE) does not approve the plant-specific implementation of ASME Code Case N-921.
2.0 REGULATORY EVALUATION
The pressurizer (PZR) pressure retaining welds at Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are ASME Code Class 1, whose ISIs are performed in accordance with Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME Code and applicable edition and addenda, as required by 10 CFR 50.55a(g).
The regulation in 10 CFR 50.55a(g)(4) states, in part, that components that are classified as ASME BPV Code Class 1, 2, and 3 must meet the requirements, except the design and access provisions and 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 regulation in 10 CFR 50.55a(z) states, in part, that alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the NRC. The licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) 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.
NUREG-1806 Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), August 2007 (ADAMS Package ML072830074),
summarizes the results of a 5-year study conducted by the NRC to develop the technical basis for revision of the Pressurized Thermal-Shock (PTS) Rule, as set forth in 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, consistent with the NRC's current guidelines on risk-informed regulation.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for SNC to request the alternative and the NRC staff to authorize it.
3.0 TECHNICAL EVALUATION
3.1 SNCs Alternative Request GEN-ISI-ALT-2024-03 Applicable ASME Code Edition Farley, Units 1 and 2, are currently in their fifth 10-year ISI interval (December 1, 2017, to November 30, 2027). Vogtle, Units 1 and 2, are currently in their fourth 10-year ISI interval (May 31, 2017, to May 30, 2027). The ASME Code of record for all four units in its respective intervals is the 2007 Edition with 2008 Addenda of the ASME Code,Section XI.
ASME Code Components Affected ASME Code Class:
Section XI, Class 1 Examination Category:
B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels B-D, Full Penetration Welded Nozzles Vessels in Vessels
Item Numbers B2.11 - PZR, shell-to-head welds, circumferential B2.12 - PZR, shell-to-head welds, longitudinal B3.110 - PZR, nozzle-to-vessel welds Component IDs:
The four tables on pages E1-1 to E1-3 of the Enclosure to the submittal, as corrected by SNC Response to a Request for Additional Information (RAI)-1 in the supplemental letter, lists the component identifications (IDs) affected.
Table 1: Farley, Unit 1, ASME Components Within Scope of Proposed Alternative ASME Category ASME Item No.
Component ID Component Description B-B B2.11 ALA1-2100-4 Pressurizer (PZR)
Bottom Head to Lower Shell B-B B2.11 ALA1-2100-7 PZR Upper Shell to Top Head B-B B2.12 ALA1-2100-1 PZR Lower Shell Long Seam B-B B2.12 ALA1-2100-3 PZR Upper Shell Long Seam B-D B3.110 ALA1-2100-9 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-10 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-11 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-12 Spray Nozzle to PZR Top Head B-D B3.110 ALA1-2100-13 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-14 Surge Nozzle to PZR Top Head
Table 2: Farley, Unit 2, ASME Components Within Scope of Proposed Alternative ASME Category ASME Item No.
Component ID Component Description B-B B2.11 APR1-2100-4 PZR Bottom Head to Lower Shell B-B B2.11 APR1-2100-7 PZR Upper Shell to Top Head B-B B2.12 APR1-2100-1 PZR Lower Shell Long Seam B-B B2.12 APR1-2100-3 PZR Upper Shell Long Seam B-D B3.110 APR1-2100-9 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-10 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-11 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-12 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-13 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-14 PZR Upper Head to Nozzle Weld Table 3: Vogtle, Unit 1, ASME Components Within Scope of Proposed Alternative ASME Category ASME Item No.
Component ID Component Description B-B B2.11 11201-V6-002-W01 Upper Head to Upper Shell Weld B-B B2.11 11201-V6-002-W05 Lower Shell to Lower Head B-B B2.12 11201-V6-002-W06 Upper Shell Longitudinal Weld B-B B2.12 11201-V6-002-W09 Lower Shell Longitudinal Weld B-D B3.110 11201-V6-002-W10 Upper Head to 6" Relief Nozzle Weld B-D B3.110 11201-V6-002-W11 Upper Head to 6" Safety Nozzle Weld B-D B3.110 11201-V6-002-W12 Upper Head to 6" Safety Nozzle Weld
Component ID Component Description B-D B3.110 11201-V6-002-W13 Upper Head to 6" Safety Nozzle Weld B-D B3.110 11201-V6-002-W14 Upper Head to 4" Safety Nozzle Weld B-D B3.110 11201-V6-002-W16 14" Surge Nozzle Weld to Lower Head Table 4: Vogtle, Unit 2, ASME Components Within Scope of Proposed Alternative ASME Category ASME Item No.
Component ID Component Description B-B B2.11 21201-V6-002-W01 Upper Head to Upper Shell Weld B-B B2.11 21201-V6-002-W05 Lower Shell to Lower Head Weld B-B B2.12 21201-V6-002-W06 Upper Shell Longitudinal Weld B-B B2.12 21201-V6-002-W09 Lower Shell Longitudinal Weld B-D B3.110 21201-V6-002-W10 Upper Head to 6" Safety Nozzle Weld B-D B3.110 21201-V6-002-W11 Upper Head to 6" Safety Nozzle Weld B-D B3.110 21201-V6-002-W12 Upper Head to 6" Safety Nozzle Weld B-D B3.110 21201-V6-002-W13 Upper Head to 6" Safety Nozzle Weld B-D B3.110 21201-V6 -002-W14 Upper Head to 4" Safety Nozzle Weld B-D B3.110 21201-V6 -002-W16 14" Surge Nozzle Weld to Lower Head Applicable ASME Code Requirements For ASME Code Class 1 welds in the PZR listed below, the ISI requirements are those specified in Subarticle IWB-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric examinations as specified in ASME Code,Section XI, Table IWB-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval:
Examination Category B-B, Item No. B2.11, PZR, shell-to-head welds, circumferential Examination Category B-B, Item No. B2.12, PZR, shell-to-head welds, longitudinal Examination Category B-D, Item No. B3.110, PZR, nozzle-to-vessel welds
Reason for Proposed Alternative In Section 4.0 of the Enclosure to its submittal dated July 3, 2024, SNC stated that the EPRI performed an assessment in the following technical report (non-proprietary) of the basis for the ASME Code,Section XI examination requirements for the PZR welds for the requested alternative.
EPRI Technical Report 3002015905, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, 2019 (hereafter referred to as EPRI Report 3002015905, ML21021A271).
The assessment includes a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). SNC stated that this report was developed consistent with EPRIs white paper on PFM (ML19241A545) and that the report concluded that the current ASME Code,Section XI ISI interval of 10 years can be increased significantly with no impact to plant safety. Based on the conclusions of the EPRI report, SNC is requesting an alternative to the ISI interval for the subject welds.
The NRC staff notes that the EPRI report was not submitted or reviewed as a topical report. The NRC staff reviewed the proposed alternative request for Farley, Units 1 and 2, and Vogtle, Units 1 and 2 as a plant-specific alternative. The NRC did not review the EPRI report for generic use, and this review does not extend beyond the Farley, Units 1 and 2, and Vogtle, Units 1 and 2 plant-specific authorization.
Proposed Alternative and Basis for Use In Section 5.0 of the Enclosure to its submittal dated July 3, 2024, SNC stated that the proposed alternative would authorize increases the ISI interval for these examination items from the current ASME Code,Section XI, requirement, to perform the examinations every other ISI interval. The licensee stated that all exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.
Duration of Proposed Alternative SNC requested to apply the proposed alternative to perform examinations every other ISI interval from the last examination performed for each item. For Farley, Units 1 and 2, this alternative covers the remainder of the fifth 10-year ISI interval through the sixth ISI interval, which is currently scheduled to end on November 30, 2037; however, Farley, Unit 1, renewed facility operating license currently expires on June 25, 2037. For Vogtle Units 1 and 2, this alternative covers the remainder of the fourth 10-year ISI interval through the fifth ISI interval, which is currently scheduled to end on May 30, 2037.
The NRC staff noted that the sixth ISI interval for Farley Units 1 is scheduled to end on November 30, 2037, which is after the expiration of the operating license on June 25, 2037. As noted in the supplemental letter dated November 15, 2024, SNC stated that the alternative requests to not perform examinations as normally scheduled during the fifth interval third period, sixth interval first period, and sixth interval second period. Examinations will resume in accordance with the ASME Section XI Code in the sixth interval third period for Farley, Unit 1,
which begins at the latest on December 1, 2035, which is prior to the Farley Unit 1 renewed facility operating license expiration at midnight on June 25, 2037.
Basis for Proposed Alternative In Section 5.0 of the Enclosure to its submittal dated July 3, 2024, SNC discussed the key aspects of the technical basis in the EPRI report and its applicability to Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The EPRI Report 3002015905 was used as basis for proposed alternative for the PZR ASME Code Examination Categories B-B and B-D welds.
The NRC staffs review focused on evaluating the applicability of the PFM analyses in Section 8.3 of EPRI Report 3002015905 and verifying whether the DFM and PFM analyses in the report support the proposed alternative. SNC cited an NRC-approved precedent for its request that were based on EPRI Report 3002015905. Specifically, the precedent of the Salem Generating Station, Units 1 and 2, submittal (ML20218A587, hereafter Salem submittal). The licensee referenced applicable portions of the technical arguments from these submittals. The NRC staff documented its review of these applications in the associated plant-specific SE (ML21145A189). The NRC staff notes that the Salem precedent discussed the Item B3.110 PZR nozzles in places but only directly addressed Items B2.11 and B2.12. For the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, the NRC staff considered the information referenced and focused on the plant-specific application of the EPRI report for Farley, Units 1 and 2, and Vogtle Units, 1 and 2.
Consistent with the key principles of the NRC risk-informed approach for performing reviews, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.
4.0 NRC STAFF EVALUATION 4.1 Degradation Mechanism In Section 5.0 of the Enclosure to its submittal dated July 3, 2024, SNC stated, in part, that:
The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the pressurizer welds covered in this request The fatigue-related mechanisms were considered in the PFM and DFM evaluations in References [9.1].
The NRC staff reviewed the submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to Farley Units, 1 and 2, and Vogtle, Units 1 and 2, to merit additional consideration. Such circumstances pertain to materials of the subject components, stress states, and reactor coolant environments. The NRC staff found that the degradation mechanisms described by the licensee for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are addressed in a manner sufficient for the applicability of EPRI Report 3002015905 and that no unknown degradation mechanisms were identified.
Specifically, the NRC staff confirmed the materials, stress states, and chemical environment
(i.e., reactor coolant) of the subject PZR welds and found them to be consistent with the assumptions and analysis made in the EPRI report and that consideration of additional degradation mechanisms is not necessary.
4.2 PFM Analysis In Section 5.0 of the Enclosure to its submittal dated July 3, 2025, SNC stated, in part, that:
Finite element analysis (FEA) was performed in Reference [9.1] to determine the stresses in the pressurizer welds covered in this request. The analysis was performed using representative Westinghouse plant geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to the SNC [licensee] PWR plants is demonstrated in Attachments 1 and 2 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] stress analysis are compared to those of the SNC PWR plants in Tables 1 and 2 [of the Enclosure].
The licensee also stated, in part, that:
Flaw tolerance evaluations were performed in Reference [9.1] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. Since the configuration considered in Reference [9.1] is consistent with the Westinghouse pressurizer design, the results of the flaw tolerance evaluation are applicable to the SNC PWR plants. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent in-service inspections (ISI), the U.S.
Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year is met.
In Section 7.0 of its submittal dated July 3, 2024, the licensee cited a number of precedents including a similar request for Salem Generating Station (Salem), Units 1 and 2. The NRC staff notes that the Salem safety evaluation authorized relief for Item Nos. B2.11 and B2.12 but did not authorize relief for the Item No. B3.110 PZR nozzle-to-vessel welds. The NRC confirmed that the plant-specific analysis provided by SNC for the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, is consistent with the approach taken in the Salem submittal for Examination Category B-B, B2.11 and B2.12 and in the Virgil C. Summer (ML23062A729), Unit 1, and Byron Station (Byron), Unit 1 and 2, and Braidwood Station (Braidwood) (ML22307A246), Units 1 and 2, submittals for Examination Category B-D, B3.110. The NRC staff determined that the PFM analysis is consistent with the authorized precedents and, therefore, finds the proposed PFM analysis to be appropriate for this application for Farley Units, 1 and 2, and Vogtle, Units 1 and
- 2. The NRC staff evaluated the licensees claim about the impact of PSI on the PFM analysis in Section 4.10 of this SE.
The NRC staff noted that the acceptance criterion of 1x10-6 failures per year (also termed Probability of Failure, PoF) is tied to that used by the NRC staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events, and other similar reviews. In that rule, the reactor vessel through-wall crack frequency (TWCF) of 1x10-6 per year for a pressurized thermal shock event is an acceptable criterion because reactor vessel TWCF is conservatively assumed to be equivalent to an increase in core damage frequency (CDF), and as such, would meet the guidance in
Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256). The discussion of TWCF is explained in detail in the technical basis document for 10 CFR 50.61a, NUREG-1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), August 2007 (ADAMS Package ML072830074).
The NRC staff also noted that the TWCF criterion of 1x10-6 per year was generated using a conservative model for reactor vessel cracking. In addition, this criterion exists within a context of reactor pressure vessel surveillance programs and inspection programs. The NRC staff finds that the licensees use of 1x10-6 failures per year that is based on the reactor vessel TWCF criterion is reasonable for the requested PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2, because (a) the impact of a PZR vessel failure would be less than the impact of a reactor vessel failure on overall risk (i.e., meaning the analyzed contribution to core damage frequency due to a PZR vessel failure would be less than the analyzed contribution due to a reactor vessel failure); (b) the subject welds have substantive, relevant, and continuing inspection histories and programs; and (c) the estimated risks associated with the individual welds are lower than the system risk criterion (i.e., the system risk is dominated by a small sub-population which can be considered the principal system risk for integrity).The failure of an individual weld is likely to represent only a limited contribution to risk). The NRC staff further noted that comparing the probability of leakage to the same criterion of 1x10-6 failures per year is conservative because leakage is not failure. The use of a PoF criteria such as 1x10-6 per year for individual welds may not be appropriate generically, but based on the discussion above, the NRC staff finds the application of this criterion acceptable for this plant-specific review for the PZR welds for Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Based on the above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for the Farley, Units 1 and 2, and Vogtle, Units 1 and 2 plant-specific alternative request.
4.3 Parameters Most Significant to PFM Results In the following sections, the NRC staff reviewed the following parameters or aspects most significant to the PFM analysis: stress analysis, fracture toughness, flaw density, flaw crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage.
4.4 Stress Analysis 4.4.1 Selection of Components and Materials In Attachment 1 and Attachment 2 to its submittal dated July 3, 2024, SNC evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI Report 3002015905 to the subject PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2, respectively. This EPRI report evaluated representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The report also specified plant-specific applicability criteria with regards to component geometries, materials, and loading conditions, that must be evaluated and met by each plant to determine the applicability of the report. The licensee stated that the plant-specific applicability of these requirements were met and that the results and conclusions of the EPRI report are applicable to Farley Units 1 and 2, and Vogtle, Units 1 and 2. The acceptability of meeting these criteria,
however, depends on the acceptability of the component and material selection described in the EPRI report, which the NRC staff evaluated below with respect to Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC staff independently evaluated the loading conditions (i.e.,
transient selection) criteria in Section 4.4.2 of this SE.
In Section 4 of EPRI Report 3002015905, EPRI discussed the variation among PZR designs.
EPRI used this information for finite element analyses (FEA, see Section 4.4.4 of this SE) to determine stresses in the analyzed components, which the licensee referenced for the corresponding PZR components requested for Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
In selecting the components, EPRI considered geometry, operating characteristics, materials, field experience with respect to service-induced cracking, and the availability and quality of component-specific information.
The NRC staff reviewed Section 4 of EPRI Report 3002015905 and finds that the PZR configurations selected in the report for stress analysis are acceptable representatives for the corresponding PZR components requested for the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, plant-specific alternative request. Specifically, the radius-to-thickness (R/t) ratios of the requested Farley, Units 1 and 2, and Vogtle, Units 1 and 2, components, provided in Tables 1 and 2 of the Enclosures to the submittal dated July 3, 2024, are either bounded by the R/t ratios analyzed in the EPRI report or are identical (e.g., the Vogtle, Units 1 and 2, PZR shell welds) to that used in the EPRI report. The NRC staff noted that Table 1 is the comparison of R/t ratios for the shell/head welds (Item Nos. B2.11 and B2.12), and Table 2 is the comparison of R/t ratios for the nozzle-to-vessel welds (Item No. B3.110). To verify the dominance of the R/t ratio, the NRC staff reviewed the through-wall stress distributions in Section 7 of the EPRI report to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. Accordingly, the NRC staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, plant-specific alternative request.
Section 9.4 of EPRI Report 3002015905 addresses criteria for plant-specific applicability of the analysis and indicates that materials are acceptable if they conform to ASME BPV Code,Section XI, Nonmandatory Appendix G, paragraph G-2110. The licensee addressed these criteria in Table 1-1 of Attachment 1 and Table 2-1 of Attachment 2 to the submittal dated July 3, 2024, for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, respectively.
The materials of the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, PZRs relevant to the requested welds are SA-533 Grade A Class 2 and SA-508 Class 2/2a. The NRC staff verified that these materials conform with ASME Code,Section XI, Paragraph G-2110. Therefore, the NRC staff finds that the materials for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, meet the material applicability criterion.
Table 1-1 of Attachment 1 and Table 2-1 of Attachment 2 to the submittal dated July 3, 2024, also state that the Farley, Unit 1 and 2, and Vogtle, Unit 1 and 2, PZR shell and nozzles meet the applicability criteria in EPRI Report 3002015905 regarding weld and nozzle configuration, attached piping line size, and thermal sleeve attachment. The NRC staff reviewed SNCs information against the applicability criteria and finds that the Farley, Unit 1 and 2, and Vogtle, Unit 1 and 2, PZRs meet the applicability criteria described in the EPRI report.
Based on the above, the NRC staff finds that the licensee has made a plant-specific case that Farley, Units 1 and 2, and Vogtle, Units 1 and 2, meet the component geometry and materials applicability criteria in the EPRI report. The analyzed geometries and materials are acceptable
for the requested PZR components at Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC concludes that the analyses acceptably model the subject Farley, Unit 1 and 2, and Vogtle, Unit 1 and 2, geometries and materials.
4.4.2 Selection of Transients In Attachment 1 and Attachment 2 to its submittal dated July 3, 2024, SNC evaluated the plant-specific applicability of the transients selected and analyzed in EPRI Report 3002015905 to the subject PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The licensee stated that the plant-specific applicability criteria regarding transients were met. The acceptability of meeting the criteria, however, depends on the acceptability of the transient selection described in the EPRI report, which the NRC staff evaluated below.
In Section 5.2 of EPRI Report 3002015905, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to PZRs. EPRI developed a list of transients for analysis applicable to the PZRs analyzed in the report, based on transients that have the largest temperature and pressure variations.
The NRC staff evaluated the transient selection in the EPRI report in detail. The NRC staff confirmed that the applicable aspects of the transients for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are addressed sufficiently. The NRC staff reviewed the discussion of transients in Section 5.2 of EPRI Report 3002015905 and determined that the transient selection defined in the report is reasonable for the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG that are expected to occur in PWRs.
In Table 1-3 and 1-4 of Attachment 1 and Table 2-3, and 2-4 of Attachment 2 to the submittal dated July 3, 2024, SNC evaluated the plant-specific applicability of the transients selected in EPRI Report 3002015905 to the PZRs of Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC staff reviewed the transient tables in the submittal dated July 3, 2024, as supplemented by letter dated November 15, 2024, and confirmed that Farley, Units 1 and 2, and Vogtle Units, 1 and 2, are bounded by the criteria in the EPRI report.
In the analyses in the EPRI report there were no separate test conditions included in the transient selection. The licensee stated in Section 5.0 of Attachment 1 to the submittal dated July 3, 2024, that pressure tests for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are performed at normal operating conditions and no hydrostatic testing has been performed since the plant began operation. The NRC staff noted that since the pressure tests are performed at normal operating conditions, they are part of Heatup/Cooldown, and, therefore, test conditions need not be analyzed as a separate transient.
Based on the above, the NRC staff finds that Farley, Units 1 and 2, and Vogtle, Units 1 and 2, meet the transient applicability criteria in the EPRI report. The analyzed transient loads are acceptable for the requested PZR components at Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC finds that the analyses acceptably model transients.
4.4.3 Other Operating Loads The NRC staff reviewed the application with regards to weld residual stress and clad residual stress. Weld residual stress and cladding stresses are addressed in EPRI Report 3002015905.
The NRC staff determined that no Farley, Units 1 and 2, or Vogtle, Units 1 and 2, plant-specific
aspects of this submittal warranted consideration because of (1) the relatively low sensitivity of the EPRI results on residual stress (Table 8-14 of EPRI Report 3002015905) and sensitivity studies conducted on stress; and (2) the small impact of clad residual stress on the PFM results.
Based on this, the NRC staff finds that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on the probability of leakage or rupture beyond the studies documented in the EPRI report.
Based on the above, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC concludes the analyses acceptably bound other operating loads.
4.4.4 Finite Element Analysis The NRC staff reviewed the application with regards to FEA. FEA were conducted in EPRI Report 3002015905 as part of the stress analysis portion of the PFM analyses. The NRC staff determined that no Farley, Units 1 and 2, or Vogtle, Units 1 and 2, plant-specific aspects of this application warranted review because the FEA were performed with the representative component geometries, materials, and loading conditions discussed in Sections 4.4.1 and 4.4.2 of this SE for which the licensee provided plant-specific information and met the plant-specific criteria.
Based on the above, the NRC staff finds that the plant-specific Farley, Units 1 and 2, and Vogtle, Units 1 and 2, submittal is acceptable with regards to FEA.
4.5 Fracture Toughness In Attachment 1 and Attachment 2 to its submittal dated July 3, 2024, SNC stated that the materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, components conform to the requirements of ASME Code,Section XI, Paragraph G-2110. As discussed in Section 4.4.1 of this SE, the NRC staff independently verified that these materials conformed to the requirements of ASME Code,Section XI, Paragraph G-2110. In EPRI Report 3002015905, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC (fracture toughness) value of 200 ksiin based on the upper-shelf fracture toughness value in the ASME Code,Section XI, A-4200. The A-4200 fracture toughness curve refers to the same fracture toughness curve in ASME Code,Section XI, Paragraph G-2110. The NRC staff also independently verified that the RTNDT value of 60°F discussed in NUREG-0800 - Chapter 5, Branch Technical Position 5-3, Revision 2, Fracture Toughness Requirements (ML070850035) bounds or is consistent with the values provided in Attachment 1 and 2 of the submittal dated July 3, 2024. The NRC staff determined that the plant-specific Farley, Units 1 and 2, and Vogtle, Units 1 and 2, submittal is acceptable with regards to fracture toughness because the materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, PZR components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.
4.6 Flaw Density In the Enclosure to its submittal dated July 3, 2024, SNC stated that, per the Salem submittal SE, a shell weld flaw density of 1.0 flaws per weld was used with a stress multiplier of 2.1 and that the probabilities of leak and rupture were still significantly below the acceptance criterion of 1x10-6 failures per year. The NRC staff noted that Section 8.3.2.2 and Table 8-9 of EPRI Report 3002015905 stated that a flaw density of 1.0 flaws per weld was used in the PZR analysis with a stress multiplier of 1.8. The NRC staff noted that, so long as the component
geometries and materials applicability criteria discussed in Section 4.4.1 of this SE are met, the use of a flaw density of 1.0 flaws per weld is sufficient for the analysis of the subject PZR welds.
As discussed in Section 4.4.1 of this SE, the licensee provided plant-specific information regarding the geometries and materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, components and met the applicability criteria. Based on the above, the NRC staff finds that the appropriate flaw density for B2.11, B2.12, and B3.110 have been considered and is, therefore, acceptable for the requested PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
4.7 ISI Schedule and Examination Coverage In Attachment 1 and Attachment 2 to its submittal dated July 3, 2024, SNC provided information on the inspection history of the requested PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2, which consists of the ISI schedule and examination coverage. The licensee stated in Section 5.0 of Attachment 1 to the submittal dated July 3, 2024, that for the Farley Units 1 and 2, PZRs, preservice inspections (PSI) have been performed followed by ISI examinations over at least four complete 10-year ISI intervals. For the Vogtle, Units 1 and 2, PZRs, PSI have been performed followed by ISI examinations over at least three complete 10-year ISI intervals.
SNC provided the inspection history of the last two completed inspections for each requested PZR welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2, in Table 1-5 and Table 1-6 of to the submittal and Table 2-5 and Table 2-6 of Attachment 2 to the submittal dated July 3, 2024. The inspection history shows that examinations for the subject components were performed during the third, fourth, and, to a limited extent, in the fifth ISI intervals for Farley, Units 1 and 2, and during the second, third, and, to a limited extent, in the fourth ISI intervals for Vogtle, Units 1 and 2. The inspection history also shows that there is no apparent evidence of unacceptable flaws in these components, which is consistent with other known operating experience and histories as summarized in Attachment 3 to the submittal dated July 3, 2024, and discussed more generally in Section 3.2 and Table 3-1 through 3-5 of EPRI Report 3002015905.
The NRC staff determined that the licensees use of only the more recent examination coverage to be reasonable because, for PZR welds, probability of rupture is relatively insensitive to examination coverage and are either similar across the two recent intervals or generally show improved coverage in the later interval. The inspection history shows that some of the examination coverages did not meet the ASME Code,Section XI, examination coverage requirement of 90 percent or greater. The licensee listed approved relief requests made under 10 CFR 50.55a(g)(5)(iii) for ASME Code,Section XI for examination coverages that were less than 90 percent as indicated in Tables 1-5 and 1-6 of Attachment 1 to the submittal, and Tables 2-5 and 2-6 of the Attachment 2 to the submittal dated July 3, 2024.
The NRC staff noted instances in which there was extremely low examination coverage (15 percent) during the ISIs for the two B3.110 welds between the 14 surge nozzle and the lower head (component IDs 11201-V6-002-W16 and 21201-V6-002-W16) at Vogtle, Units 1 and 2. In to the submittal dated July 3, 2024, SNC provided an evaluation of the impact of reduced inspection coverage on the probability of rupture and leakage for the affected welds by performing a sensitivity study on inspection coverage for the critical Case ID for the pressurizer welds (PRSHC-BW-2C) analyzed in EPRI Report 3002015905.
The licensee reported that, consistent with EPRI Report 3002015905, the sensitivity study in assumed:
(1)
A value of 1.0 flaws per weld (2)
A mean fracture toughness value of 200 ksiin and a standard deviation of 5 ksiin (3)
A stress multiplier of 1.0 (4)
The plant-specific inspection history of Farley Units 1 and 2 (PSI+10+29+30+40+70) and Vogtle Units 1 and 2 (PSI+10+20+30+60) as well as the ASME Code base case scenario (PSI+10+20+30+40+50+67+70)
(5) 100% inspection coverage was achieved for PSI consistent with the assumption and discussion in Section 8.3.5 of EPRI Report 3002015905 (6)
ISI at all plants is performed using ASME Code Section XI and therefore the Probability of Detection (POD) curve used in EPRI Report 3002015905 is applicable (7)
The POD curve for ISI can be conservatively used for PSI as discussed in to a previous RAI response from the licensee to the NRC regarding a similar alternative applicable to the steam generators at Vogtle Units 1 and 2 (ML20329A302)
(8)
Case ID PRSCH-BW-2C is the critical case and is bounding to the other cases analyzed in EPRI Report 3002015905 SNC provided that, as summarized in Table 4-2 of Attachment 4 to the submittal dated July 3, 2024, the probability of rupture for the licensees plant-specific scenarios with ISI coverage of 15 percent is identical to the probability of rupture in the ASME Code scenario and, in all cases, is approximately three orders of magnitude less than the acceptance criterion of 1x10-6 per year. The probability of leakage for the licensees plant-specific scenarios with ISI coverage of 15 percent is close to that in the ASME Code scenario and both are approximately 1.9x10-6 per year. Although this is above the acceptance criterion of 1x10-6 per year, the NRC staff notes that probability of leakage at a location does not compromise plant safety as the licensee monitors pressure boundary leakage. Leakage is detectable by plant operators, plant procedures allow for safe plant shutdown once any leakage is detected, and the probability of rupture values are maintained well below the acceptance criterion for 80 years of operation even under this limiting PSI/ISI scenario. Even though the probability of leakage increased to a value greater than 1x10-6 per year due to limited coverage less the 100%,
leakage is not component rupture and would be managed by the plant leakage detection system. Furthermore, there is minimal difference in the probabilities of rupture and leakage between the proposed alternative ISI schedule and the ASME Code Section XI specified schedule with examination ISI coverage.
The NRC staff evaluated the analysis in Attachment 4 to the submittal dated July 3, 2024, and noted that the use of CASE ID PRSHC-BW-2C is conservatively bounding because the analysis of sensitivity to inspection coverage in Table 8-33 of EPRI Report 3002015905 shows that the B3.110 surge line nozzles show less sensitivity to reduced inspection coverage (i.e., reduced volumetric examination coverage) than the case chosen for analysis in this submittal, although the NRC staff noted that Table 8-33 only reports probability of leakage and not of rupture. The NRC staff determined that this result adequately addresses the 15 percent examination coverage for welds 11201-V6-002-W16 and 21201-V6-002-W16 at Vogtle, Units 1 and 2, because the probability of rupture was much lower than the criterion of 1x10-6 per year, and because there was no change in probability of rupture values going from 100 percent to 15 percent examination coverage. As such, the result provides reasonable assurance that, for the
plant-specific alternative request for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, the limiting case in EPRI Report 3002015905 referenced for the specified welds is not sensitive to examination coverage.
Based on the above, the NRC staff finds the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, inspection history of the subject PZR welds to be acceptable and finds that the PFM approach of EPRI Report 3002015905 sufficiently represents the requested components for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, with respect to ISI schedule and examination coverage.
4.8 Other Considerations The NRC staff reviewed the application concerning initial flaw depth and length distribution, POD, models, uncertainty, and convergence. The NRC staff noted that these other considerations of the analyses in the EPRI report do not depend on plant-specific information, as compared to component geometries, materials, and transient selection for which the licensee provided plant-specific information to ensure applicability of the analyses in the report, as discussed previously.
Initial flaw depth and length distribution do not depend on plant-specific information because the flaw distribution used was based on fabrication flaws instead of service-induced flaws.
Probability of detection, which is associated with volumetric examinations, does not depend on plant-specific information because the corresponding components in different plants are subject to the same volumetric examination requirements of the ASME Code,Section XI. The models (i.e., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in fracture mechanics analyses. Uncertainty and convergence do not depend on plant-specific information because these are part of the overall PFM analyses that were addressed in the sensitivity studies and sensitivity analyses in the EPRI report.
Since these considerations are not dependent on plant-specific information, the NRC staff finds that the plant-specific Farley, Units 1 and 2, and Vogtle, Units 1 and 2, submittal is acceptable in terms of these considerations.
4.9 PFM Results Relevant to Proposed Alternative The PFM analyses in the EPRI report investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. The PFM results relevant to the proposed alternative for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, are those resulting from an ISI schedule scenario that closely matches that of Farley, Units 1 and 2, and Vogtle, Units 1 and 2, discussed in Section 4.8 of this SE. The relevant PFM results show that the probability of rupture is below the acceptance criterion of 1x10-6 failures per year. Based on the above and the discussions in Sections 4.1 through 4.9 of this SE, the NRC staff finds that the proposed alternative for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, for the requested PZR welds would result in a PoF per year that is below the acceptance criterion of 1x10-6 failures per year.
4.10 Performance Monitoring 4.10.1 Background Performance monitoring, such as ISI programs, is a necessary component such as described by the NRC five principles of risk-informed decision making. Analyses, such as PFM, work along with performance monitoring to provide a mutually supporting and diverse basis for facility condition and maintenance that is within its licensing basis. An adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation of continued appropriateness of associated analyses, and a timely method to detect novel/unexpected degradation.
4.10.2 Farley, Units 1 and 2, and Vogtle, Units 1 and 2 Evaluation SNC provided the proposed examination schedule for the subject welds in the tables on pages E1-13 through E1-16 in the Enclosure to the submittal. The proposed alternative for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, would result in approximately 20-year spans between examinations for the welds in the submittal. The 20-year span is consistent with prior precedent where U.S. licensees have sought examination relief from prescriptive ASME Section XI requirements. In the submittal, the licensee also stated that the subject welds will be examined in the same period as the last examination, but with a two-interval inspection periodicity. As noted in Section 3.6 of this SE, the sixth 10-year ISI interval for Farley, Unit 1, is scheduled to end on November 30, 2037, while the operating license expires on June 25, 2037.
However, the third period of the sixth interval for Farley, Unit 1, begins on December 1, 2034, leaving sufficient time to complete the exams scheduled for that period prior to license expiration.
The NRC staff also noted the alternative states that scope expansion will be performed in accordance with the ASME Section XI code of record. The licensee supplemented this response by the letter dated November 15, 2024, noting that: (1) if the licensee detected an indication during the alternative examinations, the indications would be reviewed against the acceptance standards of IWB-3500 and additional exams scheduled in accordance with IWB-2430 of the ASME Section XI code of record; and (2) if industry-wide operating experience indicates a new or novel degradation mechanism in the subject PZR welds, the licensee will evaluate the operating experience for applicability to the subject PZR welds.
Prolonged periods without inspection may result in a lack of monitoring and trending capacity and provide weak basis for continued adequacy of component integrity. The NRC staff performed simulations regarding potential inspection scenarios and the likelihood that such proposals would support the necessary characteristics of adequate performance monitoring.
The NRC staff sought to understand the capacity of the proposed performance monitoring plan to detect potential novel degradation. These simulations were conducted using binomial statistics and Monte Carlo methods. Based on these simulations which encapsulate the Farley, Units 1 and 2, and Vogtle, Units 1 and 2, conditions, the NRC staff determined that the previously conducted and proposed volumetric inspections would constitute sufficient performance monitoring in concert with the other aspects of the submittal reviewed by the NRC staff. The NRC staff noted that some supporting monitoring and trending information will continue to be accrued at other facilities, spread by date of application, interval schedules, and other factors, providing further assurance that adequate monitoring and trending will continue.
These characteristics regarding performance monitoring were presented, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277; agenda and slides, respectively). The NRC staff has previously applied binomial statistics and Monte Carlo methods to augment evaluation of periods beyond 20 years as well. The methods used by the NRC staff were presented at a May 25, 2022, public meeting (ML22144A345, and ML22143A840, meeting notice and presentation respectively.
Based on the above, the NRC staff determined that inspections for the subject components could be deferred during the proposed periods because an adequate level of performance monitoring is maintained for the components and the proposed performance monitoring program in conjunction with the PFM results provide an adequate level of quality and safety.
The NRC staff noted that the subject welds will be examined in the same period as the last examination, but with a two-interval inspection periodicity.
5.0 CONCLUSION
As set forth above, the NRC staff determined that the licensees proposed alternative as discussed above for the requested components provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative GEN-ISI-ALT-2024-03 for Farley, Units 1 and 2, and Vogtle, Units 1 and 2,to increase the ASME Code,Section XI, ISI interval for the subject welds, thereby deferring the examinations for those welds every other ISI interval such that all exams will occur in the same period as the last examination, but with a two-interval inspection periodicity.
All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: S. Levitus, NRR D. Dijamco, NRR Date: April 8, 2025.
ML25044A033 NRR-028 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME JLamb Zeleznock ABuford DATE 02/12/2025 02/23/2025 02/12/2025 OFFICE NRR/DORL/LPL2-1/BC NAME MMarkley DATE 04/08/2025