ML23164A120
| ML23164A120 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/30/2023 |
| From: | Jamie Heisserer Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Devlin-Gill, Stephanie | |
| References | |
| EPID L-2022-LLR-0068, FNP-ISI-ALT-05-05, Ver 1 | |
| Download: ML23164A120 (16) | |
Text
August 30, 2023 Ms. Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - PROPOSED INSERVICE INSPECTION ALTERNATIVE FNP-ISI-ALT-05-05, VERSION 1.0, TO THE REQUIREMENTS OF THE ASME CODE (EPID L-2022-LLR-0068)
Dear Ms. Coleman:
By letter dated September 30, 2022, as supplemented by letter dated February 2, 2023, Southern Nuclear Operating Company (SNC, the licensee) submitted a request to the U.S.
Nuclear Regulatory Commission (NRC) for the use of alternatives to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code (ASME Code),Section XI, for the steam generator main feedwater nozzle-to-vessel welds and nozzle inside radius sections of the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
SNC proposed to increase the ISI interval for the requested components to 20 years, from the current ASME Code Section,Section XI requirement of 10 years. 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 reviewed the proposed relief request for Farley, Units 1 and 2, as a plant-specific alternative.
The NRC staff determined that SNCs proposed alternative in FNP-ISI-ALT-05-05, Version 1.0, to increase the ISI interval from 10 years to 20 years 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 in FNP-ISI-ALT-05-05 for Farley, Units 1 and 2, for the duration of the fifth 10-year ISI interval and the sixth 10-year ISI interval.
All other requirements of ASME Code,Section XI, for which an alternative was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
J. Coleman If you have any questions, contact John G. Lamb, Senior Project Manager, at 301-415-3100 or by e-mail to John.Lamb@nrc.gov.
Sincerely, Signed by Heisserer, Jamie on 08/30/23 Jamie M. Heisserer, Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
Safety Evaluation cc: Listserv
ML23164A120 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME JLamb KGoldstein JTsao (A)
DATE 06/12/2023 06/23/2023 06/12/2023 OFFICE NRR/DORL/DD NAME JHeisserer DATE 08/30/2023 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR ALTERNATIVE FNP-ISI-ALT-05-05, VERSION 1.0 STEAM GENERATOR MAIN FEEDWATER NOZZLE-TO-VESSEL WELDS AND NOZZLE INNER RADII ASME CODE, SECTION XI, EXAMINATION CATEGORY C-B SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated September 30, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22273A159), as supplemented by letter dated February 2, 2023 (ML23033A603), Southern Nuclear Operating Company (SNC, the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) alternative FNP-ISI-ALT-05-05, Version 1.0, which proposed an alternative to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)Code (ASME Code),
Section XI, for the steam generator (SG) main feedwater (FW) nozzle-to-shell welds (NSWs) and nozzle inside radius (NIR) sections of the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
SNC proposed to increase the ISI interval for the requested components from the current ASME Code,Section XI, requirement of 10 years. The regulation in 10 CFR 50.55a(z)(1) requires the licensee to demonstrate that the proposed alternative provides an acceptable level of quality and safety. The NRC staff reviewed the proposed alternative request for Farley units as a plant-specific alternative.
2.0 REGULATORY EVALUATION
The SG FW NSWs and NIR sections at the Farley units are ASME Code Class 2 components, 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).
Enclosure
The regulations in 10 CFR 50.55a(g)(4) state, in part, that components that are classified as ASME Code Class 1, 2, and 3 components shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a may be used when authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Proposed Alternative
Applicable Code Edition and Addenda
The Farley units are currently in the fifth 10-year ISI interval (December 1, 2017, to November 30, 2027), and the ASME Code of record for this ISI interval is the 2007 Edition (2008 Addenda) of the ASME Code,Section XI.
Components Affected ASME Code Class: Section XI, Class 2 Examination Category: C-B, Pressure Retaining Nozzle Welds in Pressure Vessels Item Numbers: C2.21 for the SG FW NSWs C2.22 for the SG FW NIR sections Component Numbers: As listed below.
Unit Component ID ASME Code Component Description Item No.
UNIT 1 - 16" MAIN FEEDWATER NOZZLE TO 1 ALA2-3100-8R C2.21 SHELL WELD UNIT 1 - 16" MAIN FEEDWATER NOZZLE TO 1 ALA2-3200-8R C2.21 SHELL WELD UNIT 1 - 16" MAIN FEEDWATER NOZZLE TO 1 ALA2-3300-8R C2.21 SHELL WELD ALA2-3100- UNIT 1 - MAIN FEEDWATER NOZZLE INNER 1 C2.22 IR8R RADIUS ALA2-3200- UNIT 1 - MAIN FEEDWATER NOZZLE INNER 1 C2.22 IR8R RADIUS ALA2-3300- UNIT 1 - MAIN FEEDWATER NOZZLE INNER 1 C2.22 IR8R RADIUS UNIT 2 - 16" MAIN FEEDWATER NOZZLE TO 2 APR2-3100-8R C2.21 SHELL WELD UNIT 2 - 16" MAIN FEEDWATER NOZZLE TO 2 APR2-3200-8R C2.21 SHELL WELD
Unit Component ID ASME Code Component Description Item No.
UNIT 2 - 16" MAIN FEEDWATER NOZZLE TO 2 APR2-3300-8R C2.21 SHELL WELD APR2-3100- UNIT 2 - MAIN FEEDWATER NOZZLE INNER 2 C2.22 IR8R RADIUS APR2-3200- UNIT 2 - MAIN FEEDWATER NOZZLE INNER 2 C2.22 IR8R RADIUS APR2-3300- UNIT 2 - MAIN FEEDWATER NOZZLE INNER 2 C2.22 IR8R RADIUS ASME Code Requirements for Which Alternative Is Requested The ASME Code Examination Category C-B, Item No. C2.21, of Table IWC-2500-1 requires a surface and volumetric examination of the SG FW NSWs at terminal ends of piping runs during each Section XI 10-year ISI interval. Examination Category C-B, Item No. C2.22, of Table IWC-2500-1 requires a volumetric examination of terminal ends of piping runs the SG FW NIR sections during each Section XI 10-year ISI interval. For both C2.21 and C2.22, the required volumes (and surfaces for C2.21) are shown in the appropriate figure in IWC-2500-4(a), (b), or (d). As noted in Table IWC-2500-1 for Examination Category C-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.
Reason for Proposed Alternative In Section 4.0 of the Enclosure to its submittal dated September 30, 2022, SNC stated the following:
The Electric Power Research Institute (EPRI) performed an assessment [1] of the basis for the ASME Section XI examination requirements specified for Examination Category C-B of ASME Section XI, Division 1 for Steam Generator (SG) Main Steam (MS) and Feedwater (FW) Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. The assessment includes a survey of inspection results from 74 units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [1] report concluded that the current ASME Code Section XI inspection interval of ten years can be increased significantly with no impact to plant safety. Based on the conclusions of the EPRI report supplemented by plant-specific evaluations contained herein, Southern Nuclear Company (SNC) is requesting an alternate inspection interval for the subject components. The Reference [1] report was developed consistent with the recommendations provided in EPRIs White Paper on PFM [11].
The referenced report is EPRI Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR [pressurized water reactors] SG Class 1 Feedwater and MS NSWs and Nozzle Inside Radius Sections, 2019 (hereinafter referred to as EPRI Report 3002014590, ML19347B107). EPRI has not submitted the report for NRC review and approval. The NRC staff reviewed the proposed alternative request for the Farley, Units 1 and 2, as a plant-specific alternative. The NRC staff did not review EPRI Report 3002014590 for generic use, and this review does not extend beyond the plant-specific authorization for the Farley, Units 1 and 2.
Proposed Alternative In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC stated that the proposed alternative is to increase the ISI interval for the affected components from the current ASME Code,Section XI requirement of 10 years to 20 years for the remainder of the fifth 10-year ISI interval and through the sixth 10-year ISI interval. The licensee also stated that all examinations 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 alternative for the remainder of the fifth 10-year ISI interval and through the sixth 10-year ISI interval of the Farley, Units 1 and 2. The fifth 10-year ISI interval began on December 1, 2017, and is currently scheduled to end on November 30, 2027. The sixth 10-year ISI interval is currently scheduled to begin on December 1, 2027, and ends on November 30, 2037.
Basis for Proposed Alternative In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC discussed the key aspects of the technical basis and its applicability to the Farley, Units 1 and 2. These key aspects contained on pages E-2 through E-8 in the submittal dated September 30, 2022, are (1) applicability of the degradation mechanism evaluation in EPRI Report 3002014590 to the Farley, Units 1 and 2; (2) applicability of the stress analysis in EPRI Report 3002014590 to the Farley, Units 1 and 2; (3) applicability of the flaw tolerance evaluation in EPRI Report 3002014590 to the Farley, Units 1 and 2, and (4) inspection history.
3.2 NRC Staff Evaluation The NRC staff reviewed SNCs technical basis for the proposed alternative pursuant to 10 CFR 50.55a(z)(1). The NRC staffs review focused on evaluating the applicability of the PFM analyses in Section 8.3 of EPRI Report 3002014590, and verifying whether the DFM and PFM analyses in the report support the proposed alternative. The licensee cited an NRC-approved precedent for its request that was based on EPRI Report 3002014590. This precedent included a Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, submittal (ML19347B105, hereafter Vogtle, Units 1 and 2, submittal). SNC referenced applicable portions of the technical arguments from the Vogtle, Units 1 and 2, submittal. The NRC staff documented its review of the Vogtle, Units 1 and 2, submittal in the associated plant-specific safety evaluation (SE) for Vogtle (ML20352A155). For the Farley, Units 1 and 2, review, the NRC staff considered the information referenced and focused on the plant-specific application of EPRI Report 3002014590 for the Farley, Units 1 and 2. Consistent with the key principles of the NRC risk-informed approach, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.
The NRC staff did not review EPRI Report 3002014590 for generic use, and this review does not extend beyond the plant-specific authorization for the Farley, Units 1 and 2.
3.2.1 Degradation Mechanisms In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC stated, in part, that:
An evaluation of degradation mechanisms that could potentially impact the reliability of the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections was performed for the industry in Reference [1] - [ML19347B105].
Evaluated mechanisms 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 SG MS and FW nozzles. This observation was acknowledged by the NRC in Section 3.8, page 6, second paragraph of the Reference [12] Safety Evaluation (SE) [ML20352A155] for the Vogtle Units 1 & 2 Request for Alternative. Since the materials and operating conditions of the SG feedwater nozzles considered in Reference [1] are consistent with Farley Units 1
& 2 (per Tables A1 and A2 of Appendix A), the Reference 1 conclusion is also applicable to Farley Units 1 & 2. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in Reference [1].
The NRC staff reviewed the submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to the Farley, Units 1 and 2, to merit additional consideration, such circumstances pertain to materials of the subject components, stress states, and reactor coolant environment. The NRC staff finds that no circumstances at the Farley, Units 1 and 2, that would require consideration of a unique degradation mechanism beyond application of EPRI Report 3002014590. Specifically, the NRC staff reviewed the materials, stress states, and chemical environment (i.e., reactor coolant) of the subject SG FW NSWs and NIR sections of the Farley, Units 1 and 2, and found it to be consistent with the assumptions made in the EPRI report. Therefore, the NRC staff finds that consideration of additional degradation mechanisms beyond those from EPRI Report 3002014590 is not necessary.
3.2.2 Probabilistic Fracture Mechanics (PFM) Analysis In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC stated, in part, that:
Finite element analysis (FEA) was performed in Reference [1] - [ML19347B105]
to determine the stresses in the SG FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections for representative plants in the industry. The analysis 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 Farley Units 1 & 2 is shown in Appendix A and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [1] stress analysis are compared to those at Farley Units 1 & 2 in Table 1.
The licensee also stated, in part, that:
Flaw tolerance evaluations were performed in Reference [1] - [ML19347B105]
consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. 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 safety goal of 10-6 failures per year is met.
The NRC staff confirmed that the analysis provided by SNC in the submittal for the Farley, Units 1 and 2, is consistent with the approach taken in the Vogtle precedent and explicitly referenced by the licensee for plant-specific applicability in the request for the Farley, Units 1 and 2. The NRC staff determined that the PFM analysis is consistent and, therefore, finds the proposed PFM analysis to be appropriate for this application for the Farley, Units 1 and 2.
The NRC staff noted that the acceptance criterion of 10-6 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 10-6 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, 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 (ML072830074).
The NRC staff also noted that the TWCF criterion of 10-6 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 SNCs use of 10-6 failures per year that is based on the reactor vessel TWCF criterion is reasonable for the requested SG FW NSWs and NIR sections of the Farley, Units 1 and 2, because (a) the impact of an SG vessel failure would be less than the impact of a reactor vessel failure on overall risk (i.e., meaning contribution to core damage frequency due to an SG vessel failure would be less than the 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 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, which means that failure of an individual weld is likely to lead to only a limited contribution to risk). The NRC staff further noted that comparing the probability of leakage to the same criterion of 10-6 failures per year is conservative, because leakage is not failure. The use of a PoF criteria, such as 10-6 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 SG FW NSWs and NIR sections of the Farley, 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 plant-specific alternative request for the Farley, Units 1 and 2.
3.2.3 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, fatigue crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage.
3.2.4 Stress Analysis 3.2.4.1 Selection of Components and Materials In Appendix A to the Enclosure to its submittal dated September 30, 2022, SNC evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI Report 3002014590 (ML19347B105) to the subject SG FW NSWs and NIR sections of the Farley, Units 1 and 2. This report evaluated representative SG 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 SG 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 criteria regarding component geometries and materials of the Farley, Units 1 and 2, were met. The acceptability of meeting these criteria, however, depends on the acceptability of the component and material selection described in EPRI Report 3002014590, which the NRC staff evaluated below. The NRC staff evaluated the loading conditions (i.e., transient selection) criteria in Section 3.2.4.2 of this SE.
In Section 4 of EPRI Report 3002014590, EPRI discussed the variation among SG shell and SG nozzle designs. EPRI used this information for FEA (see Section 3.2.4.4 of this SE), to determine stresses in the analyzed components, which the licensee referenced for the corresponding SG components requested for the Farley, 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 3002014590 and finds that the SG configurations selected in the report for stress analysis are acceptable representatives for the corresponding SG components requested for the Farley plant-specific alternative request.
Specifically, the radius-to-thickness (R/t) ratio of the SG shell of the Farley, Units 1 and 2, is bounded by radius/thickness (R/t) ratio for the SG shell analyzed in the report, as shown in Table 1 of the Enclosure to the submittal, dated September 30, 2022. The R/t ratio of the SG FW nozzle of the Farley, Units 1 and 2, is slightly higher than the R/t ratio for the SG FW nozzle analyzed in the report. SNCs sensitivity evaluation shown in Table 2 of the Enclosure to the submittal, dated September 30, 2022, addresses the slightly higher R/t ratio of the Farley SG FW nozzle. The NRC staff evaluated this sensitivity evaluation in Section 3.2.10 of this SE. The NRC staff reviewed the through-wall stress distributions in Section 7 of Report 3002014590 to confirm that the pressure stress is dominant, which confirms the dominance of the R/t ratio.
Section 9.4 of EPRI Report 3002014590 lists criteria for plant-specific applicability of the generic analysis and indicates that materials are acceptable if they conform to ASME B&PV Code,Section XI, Nonmandatory Appendix G, paragraph G-2110. The licensee addressed these criteria in Table A1 of the Enclosure to the submittal for the Farley, Units 1 and 2.
The NRC staff verified that the materials relevant to the SG FW NSWs and NIR sections of the Farley, Units 1 and 2, conform with ASME Code Section XI, Nonmandatory Appendix G.
Therefore, the NRC staff finds that the materials for Farley meet the material applicability criterion.
Table A1 of the Enclosure to the submittal, dated September 30, 2022, also states that the SG nozzles of the Farley, Units 1 and 2, meet the applicability criteria in EPRI Report 3002014590 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 SG nozzles of the Farley, Units 1 and 2, meet the applicability criteria.
Based on the above, the NRC staff finds that SNC has made a plant-specific case that the SG FW NSWs and NIR sections of the Farley, Units 1 and 2, meet the component geometry and materials applicability criteria in EPRI Report 3002014590. Therefore, the NRC staff finds that the analyzed geometries and materials in the report are acceptable for the SG FW NSWs and NIR sections of the Farley, Units 1 and 2.
3.2.4.2 Selection of Transients In Appendix A of the Enclosure to its submittal dated September 30, 2022, SNC evaluated the plant-specific applicability of the transients selected and analyzed in EPRI Report 3002014590 to the subject SG FS NSWs and NIR sections of the Farley, Units 1 and 2. The licensee demonstrated 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 3002014590, which the NRC staff evaluated below.
In Section 5.2 of EPRI Report 3002014590, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to SGs. EPRI developed a list of transients for analysis applicable to the SG shell and SG nozzles analyzed in the report, based on transients that have the largest temperature and pressure variations.
The NRC staff evaluated the transient selection in EPRI Report 3002014590 in detail in the Vogtle, Units 1 and 2, SE. The NRC staff confirmed that the generic aspects of the transients discussed in the Vogtle, Units 1 and 2, SE apply equally to the Farley, Units 1 and 2. The NRC staff reviewed the discussion of transients in Section 5.2 of EPRI Report 3002014590, and determined that the transient selection defined in the report is reasonable for the Farley plant-specific alternative request, because the selection was based on large temperature and pressure variations that are conducive to FCG that is expected to occur in PWRs. The NRC staff then compared the generic analysis in the report to plant-specific information provided in the licensees submittal dated September 30, 2022.
In Table A2 of the Enclosure to the submittal dated September 30, 2022, SNC evaluated the plant-specific applicability of the transients selected in EPRI Report 3002014590 to the Farley, Units 1 and 2, SGs. The NRC staff reviewed Table A2 of the Enclosure to the submittal dated September 30, 2022, and confirmed that the Farley, Units 1 and 2, are bounded by the criteria in the report.
In the analyses in EPRI Report 3002014590, there was no separate test conditions included in the transient selection. The licensee stated in Section 5.0 of the Enclosure to the submittal dated September 30, 2022, that pressure tests at the Farley, Units 1 and 2, are performed at normal operating conditions and no hydrostatic testing has been performed since the
replacement of the SGs. The NRC staff noted that since the pressure tests are performed at normal operating conditions, it is part of Heatup/Cooldown at the Farley, Units 1 and 2, and, therefore, test conditions need not be analyzed as a separate transient.
Based on the discussion above, the NRC staff finds that the Farley, Units 1 and 2, meet the transient applicability criteria in EPRI Report 3002014590. Therefore, the NRC staff determined that the analyzed transient loads for the SG FW NSWs and NIR sections at the Farley, Units 1 and 2, are acceptable.
3.2.4.3 Other Operating Loads The NRC staff reviewed the submittal dated September 30, 2022, with regards to other operating loads such as weld residual stress addressed in EPRI Report 3002014590. The NRC staff documented the review of this aspect of the report in the Vogtle, Units 1 and 2, SE (ML20352A155). The NRC staff determined that no plant-specific aspects of this submittal for the Farley, Units 1 and 2, warranted additional consideration, because of the relatively low sensitivity of the EPRI results on residual stress (Table 8-12 of EPRI Report 3002014590) and the sensitivity studies conducted on stress in the report. 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 EPRI Report 3002014590.
Based on the above, the NRC staff finds the treatment of other loads acceptable for the SG FW NSWs and NIR sections at the Farley, Units 1 and 2.
3.2.4.4 Finite Element Analyses (FEA)
The NRC staff reviewed the application with regards to FEA. FEA were conducted in EPRI Report 3002014590 as part of the stress analysis portion of the PFM analyses. The NRC staff documented its review of this topic in the Vogtle, Units 1 and 2, SE. The NRC staff determined that no plant-specific aspects of this submittal for the Farley, Units 1 and 2, warranted further review, because the FEA were performed with the representative component geometries, materials, and loading conditions discussed in Sections 3.2.4.1 and 3.2.4.2 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 submittal for the Farley, Units 1 and 2, is acceptable with regards to FEA.
3.2.5 Fracture Toughness In Table A1 of the Enclosure to the submittal dated September 30, 2022, SNC stated that the materials of the Farley, Units 1 and 2, SG FW nozzle and shell conform to the requirements of ASME Code,Section XI, Paragraph G-2110. As discussed in Section 3.2.4.1 of this SE, the NRC staff verified that these components, which are relevant to the SG FW NSWs and NIR sections of the Farley, Units 1 and 2, conformed to the requirements of ASME Code,Section XI, Paragraph G-2110. In EPRI Report 3002014590, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC 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 documented the review of the A-4200 fracture toughness value in the Vogtle, Units 1 and 2, SE. The NRC staff determined that the plant-specific submittal for the Farley, Units 1 and 2, is acceptable with regards to fracture toughness because the materials of the subject
Farley, Units 1 and 2, components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.
3.2.6 Flaw Density In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC stated that, per the Vogtle, Units 1 and 2, SE, a nozzle flaw density of 0.1 flaws per nozzle should have been used, and that the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10-6 failures per year. Further discussion of this topic as it relates to EPRI Report 3002014590 is contained in the Vogtle, Units 1 and 2, SE. The NRC staff noted that the flaw density of 0.1 flaws per nozzle is applicable to a specific plant so long as the component geometries and materials applicability criteria discussed in Section 3.2.4.1 of this SE are met. As discussed in that section of this SE, the licensee provided plant-specific information regarding the geometries and materials of the subject Farley, Units 1 and 2, components and met the applicability criteria. Based on the above, the NRC staff finds that the appropriate flaw density has been considered and is, therefore, acceptable for the SG FW NSWs and NIR sections of the Farley, Units 1 and 2.
3.2.7 Fatigue Crack Growth (FCG) Rate The NRC staff reviewed the application with regards to FCG rate. The FCG rate used in EPRI Report 3002014590 is based on the ASME Code,Section XI, A-4300 FCG rate. The NRC staff documented its review in detail in the Vogtle, Units 1 and 2, SE. The NRC staff noted that FCG rate depends on component material and reactor coolant environment. As discussed in Section 3.2.4.1 of this SE, the licensee provided plant-specific information regarding the materials of the subject Farley, Units 1 and 2, SG components and met the criteria for component materials. Per the ASME Code,Section XI, the A-4300 FCG rate may be used for light-water-cooled plants. Since the Farley, Units 1 and 2, are PWRs, and the PWR design is one of the two major light-water-cooled reactor designs, the NRC staff determined that the A-4300 FCG rate is appropriate for the Farley, Units 1 and 2. Based on the above, the NRC staff finds that the plant-specific submittal for the Farley, Units 1 and 2, is acceptable with regards to FCG rate.
3.2.8 Inservice Inspection (ISI) Schedule and Examination Coverage In Appendix B of the Enclosure to its submittal dated September 30, 2022, SNC provided information on the inspection history of the requested SG FW NSWs and NIR sections of the Farley, Units 1 and 2, which consists of the ISI schedule and examination coverage. In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, the licensee stated that all three SGs at Farley, Unit 1, were replaced in 2000 and that all three SGs at Farley, Unit 2, were replaced in 2001. For the replacement SGs of both Farley, Units 1 and 2, the licensee stated that PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacements. Given the implementation of ISI in the PFM analyses in EPRI Report 3002014590, the NRC staff noted that in terms of PFM modeling, ISIs with replacement would be at least as good as ISIs only because replacement is essentially repair of a postulated flaw, while the outcomes of ISI are either repair of a postulated flaw or non-detection and growth of a postulated flaw.
The inspection history (post SG replacement) of the subject SG FW NSWs and NIR sections shows that examinations up to the fourth ISI interval (up to the fifth ISI interval for one weld) were performed for both Farley, Units 1 and 2. The inspection history also shows that there is
no evidence of unacceptable flaws in these components, which is consistent with known operating histories of SGs in other nuclear plants. Finally, the inspection history shows that the examination coverages are 90 percent or greater. Based on the Farley, Units 1 and 2, inspection history, the NRC staff finds that the ISI scenarios considered in the PFM analyses in EPRI Report 3002014590 adequately represent those for the SG FW NSWs and NIR sections of the Farley, Units 1 and 2, with respect to ISI schedule and examination coverage.
3.2.9 Other Considerations The NRC staff reviewed the submittal 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 EPRI Report 3002014590 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 Report 3002014590.
The NRC staff previously reviewed the applicable aspects of these considerations as used in EPRI Report 3002014590, and documented their acceptability in detail in the Vogtle, Units 1 and 2, SE. Since these considerations are not dependent on plant-specific information, the NRC staff finds that the Farley, Units 1 and 2, plant-specific submittal is acceptable in terms of these considerations.
3.2.10 PFM Results Relevant to Proposed Alternative The PFM analyses in EPRI Report 3002014590 investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. The PFM results relevant to the proposed alternative for the Farley, Units 1 and 2, are those resulting from an ISI scenario that closely matches that of the Farley, Units 1 and 2, discussed in Section 3.2.8 of this SE. The relevant PFM results show that the probability of rupture is below the acceptance criterion of 1x10-6 failures per year.
SNC performed an additional sensitivity study for the Farley, Units 1 and 2, with only the post SG replacement ISI examinations using the PFM limiting locations, FEW-P1N and FEW-3PA, analyzed in EPRI Report 3002014590. This sensitivity study also addresses the slightly higher R/t ratio of the Farley, Units 1 and 2, SG FW nozzle discussed in Section 3.2.4.1 of this SE. The results of this additional sensitivity show that the probabilities of leakage and rupture are below the acceptance criterion of 1x10-6 failures per year.
Based on the above and the discussions in Sections 3.2.1 through 3.2.9 of this SE, the NRC staff finds that the proposed alternative for the requested SG FW NSWs and NIR sections of the Farley, Units 1 and 2, would result in a PoF per year that is below the acceptance criterion of 1x10-6 failures per year.
3.2.11 Performance Monitoring Performance monitoring, such as ISI programs, is a necessary aspect for safe operation of components, 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 are within the facilitys 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 were presented, for example, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277; agenda and slides, respectively). The NRC staff has applied binomial statistics and Monte Carlo methods to augment evaluation of extended examination intervals. The methods used by the NRC staff were presented at a May 25, 2022, and April 27, 2023, public meetings (ML22143A840 and ML23114A034, respectively).
SNC stated that the proposed alternative for the Farley, Units 1 and 2, is to increase the ISI interval for the requested components from 10 years to 20 years for the remainder of the fifth 10-year ISI interval and through the sixth 10-year ISI interval, and that examinations will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity. In the supplement dated February 2, 2023, the licensee provided the approximate dates for the next examinations of the requested components during the extended interval. The NRC staff noted that the dates provided confirm that adequate performance monitoring of the SG FW NSWs and NIR sections of the Farley, Units 1 and 2, will be maintained during the extended interval.
In Section 5.0 of the Enclosure to its submittal dated September 30, 2022, SNC discussed system leakage tests as providing further assurance of safety for the proposed alternative.
However, the NRC staff noted that the visual examinations performed during system leakage tests may not directly detect the presence or extent of degradation; may not provide direct detection of aging effects prior to potential loss of structure or intended function; and do not provide sufficient validating data necessary to confirm the modeling of degradation behavior in the subject SG FW NSWs and NIR sections. The NRC staff noted that leakage tests provide complementary additional performance monitoring to the ISI examinations but would not, in isolation, be sufficient.
Prolonged periods without inspection may result in a lack of monitoring and trending capacity and provide weak basis for continued adequacy of component integrity. Consequently, the NRC staff performed a variety of 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 conditions at the Farley, Units 1 and 2, the NRC staff determined that the previously conducted inspections and proposed volumetric inspections would constitute adequate 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.
Based on the above, the NRC staff determined that increasing the ISI interval for the requested components from 10 years to 20 years for the remainder of the fifth 10-year ISI interval and through the sixth 10-year ISI interval is acceptable, because an adequate level of performance monitoring will be maintained for the SG FW NSWs and NIR sections at the Farley, Units 1 and 2, during the extended interval.
4.0 CONCLUSION
As set forth above, the NRC staff determined that SNCs proposed alternative in FNP-ISI-ALT-05-05, Version 1.0, to increase the ISI interval from 10 years to 20 years for the requested components provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative in FNP-ISI-ALT-05-05 for the Farley units for the duration of the fifth 10-year ISI interval and the sixth 10-year ISI interval.
The NRC did not review the EPRI report for generic use, and this approval does not extend beyond the Farley plant-specific authorization.
All other requirements of Section XI of the ASME Code for which an alternative was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: D. Dijamco, NRR D. Widrevitz, NRR Date: August 30, 2023