ML25122A102
| ML25122A102 | |
| Person / Time | |
|---|---|
| Site: | Farley, Vogtle  |
| Issue date: | 05/27/2025 |
| From: | Markley M Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Turner, Zachary | |
| References | |
| EPID L-2024-LLR-0041 | |
| Download: ML25122A102 (21) | |
Text
May 27, 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-2024-002 FOR STEAM GENERATOR WELDS (EPID L-2024-LLR-0041)
Dear Ms. Coleman:
By letter dated June 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24170B057), as supplemented by letter dated October 28, 2024, (ML24302A222), Southern Nuclear Operating Company (SNC, the licensee) submitted requests for U.S. Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (ISI) alternative GEN-ISI-ALT-2024-002 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 of The 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 Table IWB-2500-1 Examination Categories B-B and Table IWC-2500-1 Examination Category C-A for Item numbers B2.40, C1.10, C1.20, and C1.30 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-2024-002 to defer the ISI examinations for certain steam generator 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 alternative and concludes, as set forth in the enclosed safety evaluation, that conformance with the proposed alternative, GEN-ISI-ALT-2024-002 provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes proposed alternative GEN-ISI-ALT-2024-002 through the Sixth 10-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 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 T. 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: 2025.05.27 11:52:49 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ALTERNATIVE REQUEST GEN-ISI-ALT-2024-002 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 June 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24170B057), as supplemented by letter dated October 28, 2024 (ML24302A222), Southern Nuclear Operating Company (SNC, the licensee) submitted requests for U.S. Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (ISI) alternative GEN-ISI-ALT-2024-002 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 Table IWB-2500-1 Examination Categories B-B and Table IWC-2500-1 Examination Category C-A for Item numbers B2.40, C1.10, C1.20, and C1.30 from every ISI interval to every other interval.
Specifically, pursuant to 10 CFR 55.55a(z)(1), SNC requested that the NRC authorize Alternative Request GEN-ISI-ALT-2024-002 to defer the ISI examinations for certain steam generator (SG) 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.
2.0 REGULATORY EVALUATION
The NRC regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state, 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 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 and 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.
NUREG-1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) screening limit in the PTS Rule (10 CFR 50.61), dated 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 PTS Rule, as set forth in 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, and is used as the basis for the NRC staff's review, consistent with the NRCs 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-2024-002 Applicable ASME Code Edition and Addenda The applicable ISI interval and associated codes of record for the subject plants are summarized in Table A.
Table A: Section XI Codes of Record for Subject Plants Plant ISI Interval ASME Section XI Code Edition/Addenda Current Interval Start Date Current Interval End Date Farley, Units 1 and 2 Fifth 2007 Edition through 2008 Addenda 12/01/2017 11/30/2027 Vogtle, Units 1 and 2 Fourth 2007 Edition through 2008 Addenda 05/31/2017 05/30/2027 ASME Code Components Affected ASME Code Class:
Class 1 and 2 Examination Categories: Class 1, B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels Class 2, C-A, Pressure Retaining Welds in Pressure Vessels Item Numbers:
B2.40 for SG vessel primary side, tubesheet-to-head welds C1.10, C1.20, and C1.30 for SG vessel secondary side welds Component IDs:
The four tables below lists the component identifications (IDs) affected for each subject plant.
Table 1: Farley, Unit 1, ASME Components Within Scope of Proposed Alternative SG ASME Category ASME Item No.
Component ID Component Description A
B-B B2.40 ALA1-3100-1R Channel Head to Tubesheet A
C-A C1.10 ALA2-3100-4R Transition Cone to Lower Shell A
C-A C1.10 ALA2-3100-5R Upper Shell to Transition Cone A
C-A C1.20 ALA2-3100-6R Elliptical Head to Upper Shell A
C-A C1.30 ALA2-3100-2R Lower Shell Barrel to Upper Tubesheet B
B-B B2.40 ALA1-3200-1R Channel Head to Tubesheet B
C-A C1.10 ALA2-3200-4R Transition Cone to Lower Shell B
C-A C1.10 ALA2-3200-5R Upper Shell to Transition Cone B
C-A C1.20 ALA2-3200-6R Elliptical Head to Upper Shell B
C-A C1.30 ALA2-3200-2R Lower Shell Barrel to Upper Tubesheet C
B-B B2.40 ALA1-3300-1R Channel Head to Tubesheet C
C-A C1.10 ALA2-3300-4R Transition Cone to Lower Shell C
C-A C1.10 ALA2-3300-5R Upper Shell to Transition Cone C
C-A C1.20 ALA2-3300-6R Elliptical Head to Upper Shel C
C-A C1.30 ALA2-3300-2R Lower Shell Barrel to Upper Tubesheet
Table 2: Farley, Unit 2, ASME Components Within Scope of Proposed Alternative SG ASME Category ASME Item No.
Component ID Component Description A
B-B B2.40 APR1-3100-1R Channel Head to Tubesheet A
C-A C1.10 APR2-3100-4R Transition Cone to Lower Shell A
C-A C1.10 APR2-3100-5R Upper Shell to Transition Cone A
C-A C1.20 APR2-3100-6R Elliptical Head to Upper Shell A
C-A C1.30 APR2-3100-2R Lower Shell Barrel to Upper Tubesheet B
B-B B2.40 APR1-3200-1R Channel Head to Tubesheet B
C-A C1.10 APR2-3200-4R Transition Cone to Lower Shell B
C-A C1.10 APR2-3200-5R Upper Shell to Transition Cone B
C-A C1.20 APR2-3200-6R Elliptical Head to Upper Shell B
C-A C1.30 APR2-3200-2R Lower Shell Barrel to Upper Tubesheet C
B-B B2.40 APR1-3300-1R Channel Head to Tubesheet C
C-A C1.10 APR2-3300-4R Transition Cone to Lower Shell C
C-A C1.10 APR2-3300-5R Upper Shell to Transition Cone C
C-A C1.20 APR2-3300-6R Elliptical Head to Upper Shel C
C-A C1.30 APR2-3300-2R Lower Shell Barrel to Upper Tubesheet
Table 3: Vogtle, Unit 1, ASME Components Within Scope of Proposed Alternative SG ASME Category ASME Item No.
Component ID Component Description 1
B-B B2.40 11201-B6-001-W08 Tube Plate to Channel Head 1
C-A C1.10 11201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 1
C-A C1.10 11201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 1
C-A C1.20 11201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 1
C-A C1.30 11201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 2
B-B B2.40 11201-B6-001-W08 Tube Plate to Channel Head 2
C-A C1.10 11201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 2
C-A C1.10 11201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 2
C-A C1.20 11201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 2
C-A C1.30 11201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 3
B-B B2.40 11201-B6-001-W08 Tube Plate to Channel Head 3
C-A C1.10 11201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 3
C-A C1.10 11201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 3
C-A C1.20 11201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 3
C-A C1.30 11201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 4
B-B B2.40 11201-B6-001-W08 Tube Plate to Channel Head
4 C-A C1.10 11201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 4
C-A C1.10 11201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 4
C-A C1.20 11201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 4
C-A C1.30 11201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld Table 4: Vogtle, Unit 2, ASME Components Within Scope of Proposed Alternative SG ASME Category ASME Item No.
Component ID Component Description 1
B-B B2.40 21201-B6-001-W08 Tube Plate to Channel Head 1
C-A C1.10 21201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 1
C-A C1.10 21201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 1
C-A C1.20 21201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 1
C-A C1.30 21201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 2
B-B B2.40 21201-B6-001-W08 Tube Plate to Channel Head 2
C-A C1.10 21201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 2
C-A C1.10 21201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 2
C-A C1.20 21201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 2
C-A C1.30 21201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 3
B-B B2.40 21201-B6-001-W08 Tube Plate to Channel Head
3 C-A C1.10 21201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 3
C-A C1.10 21201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 3
C-A C1.20 21201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 3
C-A C1.30 21201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 4
B-B B2.40 21201-B6-001-W08 Tube Plate to Channel Head 4
C-A C1.10 21201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 4
C-A C1.10 21201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 4
C-A C1.20 21201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 4
C-A C1.30 21201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld Applicable ASME Code Requirements For ASME Code Class 1 welds in the SG, the ISI requirements are those specified in Subarticle IWB-2500 of the ASME BPV Code,Section XI, which requires the licensee to perform volumetric examinations as specified in Table IWB-2500-1, for the Examination Category and Item No. listed below once every ISI interval. As noted in Table IWB-2500-1 for Examination Category B-B, for cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.
Examination Category B-B, Item No. B2.40, SG Primary Side Tubesheet-to-Head Welds For ASME Code Class 2 welds in the SG, the ISI requirements are those specified in Subarticle IWC-2500 of the ASME BPV Code,Section XI, which requires the licensee to perform volumetric and surface examinations as specified in Table IWC-2500-1, for each Examination Category and Item No. listed below once every ISI interval. As noted in Table IWC-2500-1 for Examination Category C-A, for cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.
Examination Category C-A, Item No. C1.10, Shell Circumferential Welds Examination Category C-A, Item No. C1.20, Head Circumferential Welds Examination Category C-A, Item No. C1.30, Tubesheet-to-Shell Welds
Reason for Proposed Request In Section 4.0 of the submittal dated June 18, 2025, SNC stated that the Electric Power Research Institute (EPRI) performed an assessment in the following non-proprietary report of the basis for the ASME BPV Code,Section XI, examination requirements for the SG welds identified in this SE.
EPRI Technical Report 3002015906, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 2019 (hereinafter referred to as EPRI Report, ML20225A141).
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 the EPRI Report was developed consistent with EPRIs white paper on PFM (ML19241A545) and Regulatory Guide 1.245, Preparing Probabilistic Fracture Mechanics Submittals (ML21334A158). Based on the conclusions of the EPRI Report, SNC requested an alternative to the ASME BPV Code,Section XI examination requirements for the subject SG 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 the subject plants as a plant-specific alternative. The NRC did not review the EPRI Report for generic use, and this review does not extend beyond the plant-specific authorization for Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Proposed Alternative In Section 5.0 of the Enclosure to its submittal dated June 18, 2024, SNC stated that the proposed alternative would authorize changes to the examination frequency from the current ASME BPV Code,Section XI, requirement of once every ISI interval to once 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.
The proposed alternative is to defer the ISI examinations for the affected components for the Farley, Unit 1, SG from the current ASME Code,Section XI ISI interval requirement to every other interval, for the remainder of the fifth ISI interval and through the end of the current operating license expiration on June 25, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.
The proposed alternative is to defer the ISI examinations for the affected components for Farley, Unit 2, SG from the current ASME Code,Section XI ISI interval requirement to every other interval, for the remainder of the fifth ISI interval and through the sixth ISI interval, which is currently scheduled to end of November 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.
The proposed alternative is to defer the ISI examinations for the affected components for the Vogtle, Units 1 and 2, SG from the current ASME Code,Section XI ISI interval to every other interval, for the remainder of the fourth ISI interval and through the fifth ISI interval, which is
currently scheduled to end on May 30, 2037. 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 for the remainder of the current Fifth ISI interval through the end of the Sixth ISI interval for Farley, Units 1 and 2. The NRC staff noted that the Sixth ISI interval for Farley, Unit 1, is scheduled to end on November 30, 2037, which is after the expiration of the operating license on June 25, 2037. Therefore, examinations will resume in accordance with the ASME BPV Code,Section XI 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.
SNC requested to apply the proposed alternative for the current Fourth ISI interval through the end of the Sixth ISI interval for Vogtle, Units 1 and 2.
Basis for Proposed Alternative In Section 5.0 of the Enclosure to its submittal, dated June 18, 2024, the licensee discussed the key aspects of the technical basis in the EPRI Report and its plant-specific applicability to Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The EPRI Report was used as basis for the proposed alternative for the ASME BPV Code,Section XI, Examination Categories B-B and C-A SG welds. The licensee cited precedents in Section 7.0 of the Enclosure to its submittal dated June 18, 2024.
3.2
NRC Staff Evaluation
The NRC staffs review focused on evaluating the applicability of the PFM analyses in Section 8.3 of the EPRI Report and verifying whether the DFM and PFM analyses in the report support the proposed alternative. 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.
Degradation Mechanisms In Section 5.0 of the Enclosure to its submittal dated June 18, 2024, SNC stated, in part, that:
An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in [EPRI Report]. 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 known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG welds and components covered in this requestThe fatigue-related mechanisms were considered in the PFM and DFM evaluations in [EPRI Report].
The NRC staff reviewed the SNCs submittal dated June 18, 2024, as supplemented by letter dated October 28, 2024, for plant-specific circumstances that may indicate presence of a degradation mechanism and circumstances 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 SG welds, 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 the EPRI Report and that no unknown degradation mechanisms were identified.
PFM Analysis In Section 5.0 of the Enclosure to its submittal dated June 18, 2024, SNC stated, in part, that:
Finite element analyses (FEA) were performed in [EPRI Report] to determine the stresses in the SG welds and components covered in this request. The finite element models used in [EPRI Report] are consistent with the configurations at FNP [Farley Nuclear Plant] Units 1 and 2 and VEGP [Vogtle Electric Generating Plant] Units 1 and 2 and therefore no new FEA model is required for the stress analysis of these plants. The analysis in [EPRI Report] was performed using representative pressurized water reactor (PWR) 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 FNP Units 1 and 2 and VEGP Units 1 and 2 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 analyses are compared to those of FNP Units 1 and 2 and VEGP Units 1 and 2 in Table 1.
The licensee also stated, in part, that:
Flaw tolerance evaluations were performed in [EPRI Report] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a PSI [preservice Inspection] followed by subsequent ISIs, the NRCs safety goal of 1x10-6 failures per year is met.
In Section 7.0 of its submittal dated June 18, 2024, SNC cited several precedents, including a similar plant-specific request for Millstone Power Station (Millstone), Unit 2, which the NRC authorized in its safety evaluation (SE) (ML21167A355) that authorized relief for Items B2.40, C1.10, C1.20, and C1.30. The NRC staff 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 Millstone submittal for Examination Categories B-B and C-A.
The NRC staff noted that the proposed 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 events 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, and as such meets the guidance in Regulatory Guide 1.174, An Approach to for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. This assumption is conservative because a through-wall crack in the reactor vessel does not necessarily increase the likelihood
of core damage. 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 very conservative model for reactor vessel cracking. In addition, this criterion exists within the context of reactor pressure vessel surveillance programs and inspection programs. The NRC staff finds that SNCs use of 1x10-6 failures per year based on the reactor vessel TWCF criterion is acceptable for the licensees alternative request because (a) the impact of an SG weld failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject SG welds have substantive, relevant, and continuing inspection histories and programs; and (c) the estimated risks associated with the individual welds are mostly much 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).
Based on the discussion above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for SNCs plant-specific alternative request.
Parameters Most Significant to PFM Results SNCs basis for the proposed alternative relies upon the PFM analyses presented in the EPRI Report. In the following sections, the NRC staff reviewed the parameters or aspects most significant to the PFM analysis: stress analysis, fracture toughness, flaw density, fatigue crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage. The NRC staff also reviewed the plant-specific applicability of the PFM analyses presented in the EPRI Report to Farley, Units 1 and 2, and Vogtle, Units 1 and 2, which is discussed in the sections below.
Stress Analysis Selection of Components and Materials In Attachments 1 and 2 of the submittal dated June 18, 2024, SNC evaluated the plant-specific applicability of the components and materials selected and analyzed in the EPRI Report to the subject SG welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2. This EPRI Report evaluated representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The report also defined plant-specific applicability criteria related 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 further in this SE.
In Section 4 of the EPRI Report, the EPRI discussed the variation among SG shell designs.
EPRI used this information for FEAs to determine stresses in the analyzed SG welds, which SNC referenced for the corresponding SG welds requested for Farley, Units 1 and 2, and Vogtle, Units 1 and 2. In selecting the components, the 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 the EPRI Report and finds that the SG configurations selected in the report for stress analysis are acceptable representatives for the corresponding SG welds 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, SG welds, provided in Table 1 of the Enclosure to the submittal dated June 18, 2024, are bounded by the stress multiplier of 2.1 used in the licensees plant-specific PFM analysis. 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 the EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for SNCs plant-specific alternative request.
Section 9.4 of the EPRI Report 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. SNC addressed these criteria in Tables 1-1 of and 2-1 of Attachment 2 to the submittal dated June 18, 2024, for Farley, Units 1 and 2, and Vogtle, Units 1 and 2, respectively. The licensee stated that the materials of construction for the SG welds are as reported in Table B of this SE.
Table B: Materials of Construction Unit SG Component Material Farley Units 1 and 2 vessel heads SA-508 Class 3 vessel shell SA-508 Class 3 tubesheet SA-508 Class 3 Vogtle Units 1 and 2 vessel heads SA-216 Grade WCC vessel shell SA-533 Grade A, Class 2 tubesheet SA-508 Class 2a The NRC staff verified that these materials of construction reported by the licensee conform with the material property requirements of ASME BPV Code Section XI, Nonmandatory Appendix G.
SNC further clarified that SA-216 Grade WCC material for the Vogtle, Units 1 and 2, vessel heads are low alloy steel with specified minimum yield strength of 50 kilopounds per square inch (ksi). Since the specified minimum yield strength is consistent with the dataset used to develop the ASME Code,Section XI, Nonmandatory Appendix G, fracture toughness curve, the NRC staff finds that SA-216 Grade WCC is consistent with the technical basis in the EPRI Report. In addition, the NRC staff determined that the material properties of SA-216 Grade WCC is consistent with the material properties of low alloy steel used in the Millstone, Unit 2, SGs as approved and discussed in the NRC staffs SE for the Millstone, Unit 2. Therefore, the NRC staff finds that the licensees SG materials meet the material applicability criteria.
Tables 1-1 and 2-1 of the Enclosure to the submittal dated June 18, 2024, state that the SG shells meet the applicability criteria in the EPRI Report regarding weld configuration. SNC provided supplementary drawings in Figures 1-1 and 2-1 to illustrate the weld geometries. The NRC staff reviewed the licensees information against the applicability criteria and finds that the subject SGs meet the applicability criteria described in the EPRI Report as discussed above regarding the NRC staffs review of Section 4 of the EPRI Report with respect to the subject
weld geometries.
Based on the above, the NRC staff finds that SNC 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 SG welds at Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Selection of Transients In Attachment 1 and Attachment 2 to its submittal dated June 18, 2024, SNC evaluated the plant-specific applicability of the transients selected and analyzed in the EPRI Report to the subject SG 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 the EPRI Report, the EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to the SG shell welds. The EPRI developed a list of transients for analysis applicable to the SG shell welds 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 the EPRI Report and determined that the transient selection defined in the reports are 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 and expected to occur in PWRs.
In Tables 1-2 and 1-3 of Attachment 1 and Tables 2-2 and 2-3 of Attachment 2 to the submittal dated June 18, 2024, SNC evaluated the plant-specific applicability of the transients selected in the EPRI Report to the SG welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2. The NRC staff reviewed these tables and confirmed that the transient projections are bounded by the criteria in the EPRI Report. The NRC staff noted that there were minor differences in temperatures and pressures, as described in footnote 1 of Tables 1-2 and 1-3 of the licensees submittal. However, the NRC staff determined that these minor variations would not substantially impact the stress calculations underlying the FCG calculation. Furthermore, the projected 60-year cycles in Tables 1-2, 1-3, 2-2, and 2-3 of the licensees submittal are substantially below the number of cycles assumed in the EPRI analysis. Therefore, the NRC staff finds that the transients assumed in the EPRI analysis appropriately bound Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
In the analyses in the EPRI Report, there were no separate test conditions included in the transient selection. SNC stated on page E-9 of its submittal that pressure tests (i.e., system leakage tests) 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, it is part of Heatup/Cooldown, and, therefore test conditions need not be analyzed as a separate transient.
Based on the discussion 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. Therefore, the analyzed transient loads for the subject SG welds are acceptable.
Residual Stresses 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 the EPRI Report. The NRC staff determined that no Farley, Unit 1 and 2, or Vogtle, Unit 1 and 2, plant-specific aspects of this submittal warranted consideration because of (1) the relatively low sensitivity of the EPRI results on stress (Tables 8-17 through 8-20 of the EPRI Report); 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 residual stress 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 residual stresses described in this section of the SE acceptable for the requested SG welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Finite Element Analysis In its submittal dated June 18, 2024, SNC stated that the finite element models used in the EPRI Report are consistent with the configurations at Farley, Units 1 and 2, and Vogtle, Units 1 and 2, and no new FEA model is provided for the stress analysis of these plants. The licensee further stated that it uses the FEA of the EPRI Report to determine stresses for the subject welds. The NRC staff reviewed the FEA of the EPRI Report for applicability to SGs at Vogtle, Units 1 and 2, in its safety evaluation dated January11, 2021 (ML20352A155). NRC also reviewed FEA of the EPRI Report in its evaluation of pressurizer welds for Vogtle, Units 1 and 2, and Farley Units 1 and 2, in its safety evaluation dated April 8, 2025 (ML25044A033). For the current proposed alternative, the NRC staff reviewed material and component selection, transient selection, and assumed residual stresses to determine whether the FEA of the EPRI Report is applicable to the subject welds. The NRC staff determined that SNC demonstrated that the FEA of the EPRI Report is applicable to the subject welds in each of these respects.
Based on the above, the NRC staff finds that the generic FEA performed in the EPRI Report adequately represents the plant-specific Farley, Units 1 and 2, and Vogtle, Units 1 and 2, SG welds.
Fracture Toughness In Attachment 1 and Attachment 2 to its submittal dated June 18, 2024, SNC stated that the materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, welds conform to the requirements of ASME Code,Section XI, Paragraph G-2110. The NRC staff independently verified that these materials conformed to the requirements of ASME BPV Code,Section XI, Paragraph G-2110. In the EPRI Report, the EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIc (fracture toughness) value of 200 ksiin based on the KIc curve in the ASME BPV 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 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 SG welds conform to the requirements of ASME Code,Section XI, Paragraph G-2110.
Flaw Density In the Enclosure to its submittal dated June 18, 2024, SNC described a plant-specific sensitivity study assuming 1.0 flaws per weld. The licensee stated 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 of the EPRI Report stated that a flaw density of 1.0 flaw per weld was used in the SG analysis. The NRC staff noted that, so long as the component materials applicability criteria are met, the use of a flaw density of 1.0 flaws per weld is sufficient for the analysis of the subject SG welds. SNC provided plant-specific information regarding the geometries and materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, welds and met the applicability criteria. Based on the above, the NRC staff finds that the flaw density assumed in the EPRI analysis is appropriate for the requested SG welds of Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Fatigue Crack Growth Rate The NRC staff reviewed the application with regards to FCG rate and noted that the FCG rate used in the EPRI Report is based on the ASME Code,Section XI, A-4300, FCG rate. The NRC staff noted that FCG rate depends on component material and environmental conditions. Per the ASME BPV Code,Section XI (2021 Edition), the A-4300 FCG rate may be used for low alloy ferritic steels in air and reactor water environments. The licensee provided plant-specific information regarding the materials of the subject Farley, Units 1 and 2, and Vogtle, Units 1 and 2, SG welds. Since the identified materials meet the scope of A-4300, the NRC staff determined that the A-4300 FCG rate is appropriate for Farley, Units 1 and 2, and Vogtle, Units 1 and 2.
Examination History SNC provided information on the examination history of the requested SG welds in Tables 1-4 (Farley, Unit 1), 1-5 (Farley, Unit 2), 2-4 (Vogtle, Unit 1), and 2-5 (Vogtle, Unit 2) in the Enclosure to the submittal dated June 18, 2024. These tables indicate that there were no unacceptable indications found during these examinations. The licensee stated in Section 5.0 of the submittal that the Farley, Units 1 and 2, SGs have been replaced. The licensee also described a plant-specific PFM analyzing the limiting PSI/ISI scenario of PSI+10+20+50, where a PSI was performed, an inspection was performed during the first interval after the SG replacement (+10), a second inspection was performed (+20), and a proposed inspection 30 years from the second ISI (+50). This inspection scenario captures the examination history of the Farley, Units 1 and 2, replacement SGs and bounds the examination history of Vogtle, Units 1 and 2, SGs. Therefore, the NRC staff finds that the licensee sufficiently accounted for plant-specific examination history with the supplementary PFM analysis provided in Section 5.0 of the Enclosure to the submittal dated June 18, 2024.
The NRC staff notes that the examination coverages did not meet the ASME BPV Code,Section XI, examination coverage requirement of 90 percent or greater. However, licensees are required to submit a relief request under 10 CFR 50.55a(g)(5)(iii) for ASME BPV Code,Section XI examination requirements that are determined by the licensee to be impractical, which includes examination coverages that do not meet the requirement. The NRC staff also noted an examination coverage of at least 80% for all SG welds referenced in this alternative request. In the SE for Millstone, Unit 2, the NRC staff reviewed examination coverage as low as 50% and determined that the probability of rupture is still less than the acceptance criterion of 1x10-6 per year. The subject SG welds at Farley and Vogtle have all received more coverage
than the limiting case reviewed previously by the NRC staff for Millstone. Therefore, the NRC staff finds that SNCs plant-specific examination history for the subject SG welds at Farley, Units 1 and 2, and Vogtle, Units 1 and 2, with at least 80% examination coverage, is sufficiently bounded by the PFM analyses in the EPRI Report and the plant-specific PFM described in the licensees submittal.
The subject SG welds have all received more coverage than the limiting case reviewed previously by the NRC staff. Therefore, the NRC staff finds that SNCs plant-specific examination history for the subject SG welds, with at least 80% examination coverage, is adequately bound by the PFM analyses in the EPRI Report and the plant-specific PFM in the licensees submittal.
Other Considerations The NRC staff reviewed the application concerning initial flaw depth and length distribution, probability of detection, 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 EPRI Report.
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 SG welds in different plants are subject to the same volumetric examination requirements of the ASME BPV Code,Section XI. The models (e.g., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in fracture mechanics analyses as discussed in the NRC staffs safety evaluation for Millstone, Unit 2. 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.
Modeled Examination Schedule The PFM analyses in the EPRI Report investigated several ISI examination schedule scenarios, which include pre-service inspection (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 that most closely represent the proposed alternative ISI schedules for the SG welds.
As described in section titled Examination History above of this SE, SNC performed a plant-specific PFM study for PSI+10+20+50 for Farley, Units 1 and 2, and PSI+10+30+60 for Vogtle, Units 1 and 2, to investigate the proposed examination schedules. The licensee determined that the most limiting PSI/ISI scenario is PSI+10+20+50. The NRC staff notes that this scenario is conservative relative to the proposed examination schedules of once every other ISI interval.
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, the NRC staff finds that SNCs analyses adequately bound the proposed examination schedules.
Performance Monitoring Performance monitoring, such as ISI programs, is a necessary component 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.
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.
SNCs proposed alternative consists of performing the subject SG weld examinations once every other ISI interval, as opposed to the Section XI requirement of once every ISI interval. In Section 5 of the Enclosure to the submittal dated June 18, 2024, the licensee provided proposed examination windows. The proposed examination schedule provides performance data on the subject welds throughout the licensed life of the four reactors and is conservative relative to the PFM PSI+10+20+50 scenario cited by the licensee as its technical basis. The NRC independent review determined that the technical basis for the proposed examination schedule is consistent with past NRC-approved alternative requests for Vogtle, Units 1 and 2 (ML20352A155). Therefore, the NRC staff finds that the proposed alternative provides sufficient performance monitoring for the subject welds.
Future Inspections In its letter dated June 18, 2024, the licensee discusses its plans for future inspections in accordance with the ASME Section XI code of record. In Section 5 of the proposed alternative, the licensee discusses the proposed rulemaking to adopt ASME Code Case N-921 through Incorporation by Reference in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division I. Section 5 of the proposed alternative further states that scope expansion will be performed in accordance with the ASME Section XI code of record.
The NRC staff notes that scope expansion applies if the licensee discovers unexpected degradation during a performance monitoring examination. It provides a method for the licensee to investigate extent of condition, should unexpected degradation occur.
In its request for additional information dated September 26, 2024, the NRC staff requested the licensee to describe actions if indications are detected that exceed acceptance standards of the ASME Code and corrective actions to ensure structural integrity as part of demonstrating an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1). In its supplement dated October 28, 2024, SNC clarified that scope expansion would occur according to the rules of ASME BPV Code,Section XI IWB-2420, IWB-2430, IWC-2420, and/or IWC-2430. These rules provide criteria for expanding examination scope within the reactor unit where the degradation was discovered. The NRC independent review determined that SNC is proposing a 50% reduction in the examination burden (i.e., once every other interval) as compared to 75%
reduction that has been previously approved by the NRC staff. The NRC staff also notes that
the final rulemaking incorporating ASME Code Case N-921, was incorporated in Revision 1 to Regulatory Guide 1.147 with certain conditions. The NRC staff finds the licensees proposed future inspections and the scope expansion plans, in concert with the increased level of performance monitoring, is acceptable.
4.0 CONCLUSION
As set forth above, the NRC staff determined that SNCs proposed alternative GEN-ISI-ALT-2024-002 for the requested SG welds provides an acceptable level of quality and safety for the affected components in Table 1 through 4 above. Accordingly, the NRC staff concludes that SNC 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 for the remainder of the current fifth ISI interval through the end of the sixth ISI interval for Farley, Units 1 and 2, and for the remainder of the current fourth ISI interval through the end of the sixth ISI interval for Vogtle, Units 1 and 2. Please note that the proposed alternative GEN-ISI-ALT-2024-002 does not go beyond the current renewed license date of June 25, 2025, for Farley, Unit 1.
All other ASME BPV Code,Section XI requirements for which relief has not been specifically requested and approved in this alternative request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: Michael Benson, NRR John Tsao, NRR Date: May 27, 2025
ML25122A102 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME JLamb KZeleznock ABuford (JTsao for)
DATE 05/02/2025 05/05/2025 05/23/2025 OFFICE NRR/DORL/LPL2-1/BC NAME MMarkley DATE 05/27/2025