ML25105A064
| ML25105A064 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/29/2025 |
| From: | Tony Nakanishi Plant Licensing Branch IV |
| To: | Entergy Operations |
| Galvin, Dennis | |
| References | |
| EPID L-2024-LLR-0038 | |
| Download: ML25105A064 (1) | |
Text
April 29, 2025 ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.
N-ADM-8 1448 S.R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - AUTHORIZATION OF ALTERNATIVE REQUESTS ANO1-ISI-24-01 AND ANO2-ISI-24-01 REGARDING EXTENSION OF STEAM GENERATOR INSERVICE INSPECTION INTERVAL (EPID L-2024-LLR-0038)
Dear Sir or Madam:
By letter dated June 6, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24158A389), as supplemented by letters dated October 16, 2024, and February 20, 2025 (ML24290A098 and ML25051A292, respectively), Entergy Operations, Inc. (the licensee) submitted requests for alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 to the U.S. Nuclear Regulatory Commission (NRC). The licensee proposed alternatives to the inspection requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternatives in Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01 to defer the inservice inspection (ISI) of the steam generator welds and nozzle inner radii at ANO-1 and ANO-2 to the end of the operating licenses (i.e., May 20, 2034, for ANO-1; and July 17, 2038, for ANO-2).
As set forth in the enclosed safety evaluation, the NRC staff has determined that the licensees proposed alternatives would provide an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI, requirements and inspection items specified and referenced in Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01. Accordingly, the 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 Alternative Request ANO1-ISI-24-01 at ANO-1 up to May 20, 2034, and Alternative Request ANO2-ISI-24-01 at ANO-2 up to July 17, 2038.
All other ASME Code,Section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the ANO-1 Project Manager, Dennis Galvin at (301) 415-6256 or by email at Dennis.Galvin@nrc.gov.
Sincerely, Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313 and 50-368
Enclosure:
Safety Evaluation cc: Listserv TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.04.29 12:04:29 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUESTS FOR ALTERNATIVES ANO1-ISI-24-01 AND ANO2-ISI-24-01 REGARDING EXTENSION OF STEAM GENERATOR INSERVICE INSPECTION INTERVAL ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT 1 AND 2 DOCKET NO. 50-313 AND 50-368
1.0 INTRODUCTION
By letter dated June 6, 2024 (Agencywide Documents Access and Management System Accession No. ML24158A389), as supplemented by letters dated October 16, 2024, and February 20, 2025 (ML24290A098 and ML25051A292, respectively), Entergy Operations, Inc.
(the licensee), submitted requests for alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 to the U.S. Nuclear Regulatory Commission (NRC). The licensee proposed alternatives to the inspection requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, for Arkansas Nuclear One (ANO), Units 1 and 2 (ANO-1 and ANO-2, respectively).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
Acceptable level of quality and safety, the licensee requested to use the proposed alternatives in Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01 to defer the inservice inspection (ISI) of the steam generator (SG) welds and nozzle inner radii at ANO-1 and ANO-2 to the end of the operating licenses (i.e., May 20, 2034, for ANO-1; and July 17, 2038, for ANO-2).
2.0 REGULATORY EVALUATION
Regulatory Requirements The SG pressure-retaining welds and nozzles at the subject units are ASME Code Class 1 and Class 2 components, whose ISIs are performed in accordance with the applicable edition of the ASME Code,Section XI, as required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements.
Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in ASME Code,Section XI.
The regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, states, in part, that alternatives to the requirements of 10 CFR 50.55a(b) through (h) may be used, when authorized by the Director, Office of Nuclear Reactor Regulation, if (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensees Proposed Alternative 3.1.1 ASME Code Components Affected The affected components are SG pressure-retaining welds and full penetration welded nozzles (nozzle-to-shell welds and nozzle inside radius sections). The required examinations of the affected SG components are specified by the examination category and item number of the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1 as discussed below.
Examination Category:
Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 1, Category B-D, full penetration welded nozzles in vessels Class 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-B, pressure retaining nozzle welds in pressure vessels Item Numbers:
B2.31 - SG (Primary side), head welds, circumferential B2.40 - SGs (primary side), tubesheet-to-head weld B3.130 - SGs (primary side), nozzle-to-vessel welds C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell weld C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections The plant-specific identification number and description for each affected component are shown in the enclosure to the submittal dated June 6, 2024.
3.1.2 Applicable ASME Code Edition and Addenda ANO-1 and ANO-2 are in fifth 10-year ISI intervals that are scheduled to end on May 30, 2027, and March 25, 2030, respectively. The codes of record for the fifth 10-year ISI intervals are the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI. The applicable Code for the sixth 10-year ISI intervals will be selected in accordance with the requirements of 10 CFR 50.55a, Codes and standards.
3.1.3 Applicable ASME Code Requirement The ASME Code,Section XI, IWB-2500(a), table IWB-2500-1, Examination Categories B-B and B-D; and IWC-2500(a), table IWC-2500-1, Examination Categories C-A and C-B require examination of the following item numbers.:
Item No. B2.31 -- volumetric examination of all nozzles during the first ASME Code,Section XI, inspection interval and one weld per head during successive intervals. The examination areas are shown in the ASME Code,Section XI, figure IWB-2500-3.
Item No. B2.40 -- volumetric examination of essentially 100 percent of the length of all welds during the first inspection interval. For successive inspection intervals, the examination may be limited to one vessel among the group of vessels performing a similar function. The examination volume is shown in the ASME Code,Section XI, figure IWB-2500-6.
Item No. B3.130 -- volumetric examination of all nozzles during each inspection interval. The examination areas are shown in the ASME Code,Section XI, figures IWB-2500-7(a), (b), (c) and (d).
Item No. C1.20 -- volumetric examination of essentially 100 percent of the length of the head-to-shell welds during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in the ASME Code,Section XI, figure IWC-2500-1.
Item No. C1.30 -- volumetric examination of essentially 100 percent of the length of the tubesheet-to-shell welds during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in the ASME Code,Section XI, figure IWC-2500-2.
Item No. C2.21 -- volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination area and volume are shown in the ASME Code,Section XI, figures IWC-2500-4(a), (b), or (d).
Item No. C2.22 -- volumetric examination of all nozzles inside radius sections at terminal ends of piping runs during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in the ASME Code,Section XI, figures IWC-2500-4(a), (b),
or (d).
3.1.4 Reason for Request In the June 6, 2024, submittal, the licensee stated that the Electric Power Research Institute (EPRI) performed assessments in two technical reports (EPRI report 3002014590, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections (hereafter EPRI report 14590) April 2019 (ML19347B107); and EPRI 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, Tube sheet-to-Head and Tube sheet-to-Shell Welds(hereafter EPRI Report 15906), 2019 (ML20225A141)), to provide the bases for the
ASME Code,Section XI, examination requirements specified for the subject SG welds and nozzle inner radii.
The licensee further stated in its submittal that the EPRI assessments include 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). The two EPRI reports concluded that the current ASME Code,Section XI, ISI examinations can be deferred for some time with no impact to plant safety. Based on the conclusions of the two EPRI reports supplemented by plant-specific evaluations, the licensee requested to defer ISI for the subject SG components. The licensee explained that the two EPRI reports were developed consistent with the recommendations provided in EPRIs White Paper entitled, White Paper on Suggested Content for PFM Submittals to the NRC, dated February 27, 2019 (ML19241A545), and NRC Regulatory Guide (RG) 1.245, Preparing Probabilistic Fracture Mechanics Submittals, January 2022 (ML21334A158) for PFM submittals and associated technical basis NRC report NUREG/CR-7278, Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications, January 2022 (ML22014A406).
The NRC staff noted that the two EPRI reports were not submitted or reviewed as topical reports. The staff reviewed the proposed alternative request for ANO-1 and ANO-2, as plant-specific alternatives. The staff did not review the EPRI reports for generic use, and this review does not extend beyond the ANO-1 and ANO-2 plant-specific authorization.
3.1.5 Proposed Alternative ANO-1 The licensee requested an inspection alternative to the examination requirements of the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers at ANO-1:
ASME Code Category Item No.
Description B-B B2.31 SGs (primary side), head welds, circumferential B-B B2.40 SGs (primary side), tube sheet-to-head weld C-A C1.30 Tube sheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Nozzle inside radius sections The licensee replaced the ANO-1 SGs in the third period of the third ISI interval in 2005. The new SG welds and nozzle inner radii received the required preservice inspection (PSI) examination followed by ISI examinations through the second period of the current fifth ISI interval.
The licensee stated that the proposed alternative is to defer the ISI examinations of these item numbers for the ANO-1 SGs from the current 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on May 20, 2034. The licensee further stated that this equates to an extension of 16 years, 11 months, 20 days from the end of the fourth ISI interval (May 30, 2017).
ANO-2 The licensee requested an inspection alternative to the examination requirements of the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers at ANO-2:
ASME Code Category Item No.
Description B-B B2.40 SGs (primary side), tube sheet-to-head weld B-D B3.130 SGs (primary side), nozzle-to-vessel welds C-A C1.20 Head circumferential weld C-A C1.30 Tube sheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Nozzle inside radius sections The licensee replaced the ANO-2 SGs during the first period of the third ISI interval in 2000. The licensee stated that the new SG welds and nozzle inner radii received the required PSI examinations followed by ISI examinations through the first period of the current fifth inspection interval.
The NRC staff noted that table 2-6 of attachment 2 to the enclosure of the June 6, 2024, submittal, indicates that the latest inspections of ANO-2 SG welds were performed in the fourth, not fifth, ISI interval. In its supplement dated October 16, 2024, the licensee clarified that it has not performed any inspections of the SG welds in the fifth 10-year interval at ANO-2. As part of clarification, the licensee provided end dates of second, third, fourth and fifth ISI intervals at ANO-2.
The licensee stated that the proposed alternative is to defer the ISI examinations of these item numbers for the ANO-2 SGs from the current 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on July 17, 2038. The licensee explained that this equates to an extension of 17 years, 3 months, 23 days from the end of the fourth ISI interval (March 25, 2021).
3.1.6 Duration of Proposed Alternative For ANO-1, the proposed alternative is requested for the remainder of the current fifth inspection interval and through the end of currently licensed operating life, which is currently scheduled to end on May 20, 2034.
For ANO-2, the proposed alternative is requested for the remainder of the current fifth inspection interval and through the end of currently licensed operating life, which is currently scheduled to end on July 17, 2038.
3.2
NRC Staff Evaluation
As stated above, the technical basis for the licensees proposed alternatives is based on EPRI reports 14590 and 15906. The NRC staff has approved similar alternative requests from other nuclear plants that use these two EPRI reports as shown in the staffs safety evaluations (SEs) for Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle) (ML20352A155) and Millstone Power Station, Unit 2 (Millstone) (ML21167A355). The staffs SEs for Vogtle and Millstone have
identified issues related to the applicability of the EPRI reports to plant-specific applications regarding SG component examinations. Therefore, for its review efficiency, the staff evaluated the applicability of NRC-approved Vogtle and Millstone submittals and the EPRI Reports to the ANO-1 and ANO-2 alternative requests.
The PFM analysis in EPRI report 14590 uses the Probabilistic Optimization of Inspection (PROMISE) Version 1.0 software. As part of the NRC staffs review of Vogtles alternative request, the staff audited the PROMISE Version 1.0 software as discussed in the staffs audit plan (ML20128J311). The staff determined that PROMISE Version 1.0 is acceptable for use on a case-by-case and plant-specific basis. However, the staff has not reviewed the PROMISE Version 2.0 software, which is used in EPRI report 15906. The key difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. Versions 1.0 and 2.0 assume 100 percent coverage for the PSI examination.
3.2.1 Degradation Mechanisms The licensee stated that both EPRI reports evaluated degradation mechanisms that could affect the reliability of the subject SG components, including stress corrosion cracking, environmental assisted fatigue (EAF), microbiologically influenced corrosion, pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion, general corrosion, galvanic corrosion, and mechanical/thermal fatigue. The licensee further stated that other than the potential for EAF and mechanical/thermal fatigue, there were no known active degradation mechanisms identified that would significantly affect the long-term structural integrity of the subject SG components. The licensee indicated that fatigue-related mechanisms were considered in the PFM and DFM evaluations in both EPRI reports. The licensee explained that the materials and operating conditions in the proposed ANO-1 and ANO-2 alternatives are consistent with those addressed in both EPRI reports.
The NRC staff found no evidence of conditions at ANO-1 and ANO-2 that would require consideration of a unique degradation mechanism beyond application of the information the licensee referenced from the EPRI reports. Specifically, the staff reviewed the materials, stress states, and consistency of chemical environment (i.e., reactor coolant) of the subject SG components and found them to be consistent with the assumptions made in the EPRI reports.
Therefore, the NRC staff finds that consideration of additional degradation mechanisms beyond those from the EPRI reports is not necessary.
3.2.2 PFM Analysis The NRC staff noted that the acceptance criterion of 1x10-6 failures per year (referred to as probability of failure, PoF) is tied to that used by the staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events. In that rule, the reactor vessel through-wall crack frequency (TWCF) of 1x10-6 per year for a pressurized thermal shock event is an acceptable criterion because reactor vessel TWCF is conservatively assumed to be equivalent to an increase in core damage frequency, and meets the guidance in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018 (ML17317A256). The staff finds that the TWCF 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 NUREG-1806 Technical Basis for
Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), August 2007 (Package ML072830074), which provides the technical basis for 10 CFR 50.61a. The staff noted that the TWCF criterion of 1x10-6 per year was generated using a conservative model for reactor vessel cracking. In addition, this criterion exists within a context of reactor pressure vessel material surveillance programs and inspection programs.
The NRC staff finds that the licensees use of 1x10-6 failures per year based on the reactor vessel TWCF criterion is acceptable for the subject SG component examination because (a) the impact of a SG vessel failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject SG components have substantive, relevant, and continuing inspection histories and programs; and (c) the estimated risks associated with the individual weld or nozzle inside radius 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).
The NRC staff acknowledges that comparing the probability of leakage to the same TWCF criterion is conservative because component leakage is less severe than component rupture.
The staff notes that the use of a PoF criteria such as 1x10-6 per year for individual welds or nozzles may not be appropriate generically. However, based on the discussion above, the NRC staff finds the application of the TWCF criterion acceptable for the proposed examinations of SG weld and nozzle inside radius at ANO-1 and ANO-2.
The NRC staff noted that the acceptance criterion of 1x10-6 failures per year is lower (i.e.,
conservative) than the criterion the staff accepted in the following three industry reports:
(1) proprietary report BWRVIP-05, BWR [Boiling Water Reactor] Vessel and Internals Project:
BWR Reactor Pressure Vessel Weld Inspection Recommendations, September 1995; (2) non-proprietary report BWRVIP-108NP-A, BWR Vessel and Internals Project: Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, October 2018 (ML19297F806); and (3) non-proprietary report BWRVIP-241NP-A, BWR Vessel and Internals Project: Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, October 2018 (ML19297G738). These reports were developed prior to or around the time the rules for PTS were reevaluated. As such, the acceptance criterion for the failure frequency in the reports is based on the guidelines for PTS analysis in RG 1.154, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors that were available at the time. The NRC staff also noted that the BWRVIP topical reports included substantive inspection aspects that were critical to the NRCs findings.
Based on the discussion above, the NRC staff finds that the use of the acceptance criterion of 1x10-6 failures per year for PoF is acceptable for the ANO-1 and ANO-2 alternative requests.
The staff confirmed that the PFM analysis for the ANO-1 and ANO-2 alternative requests is consistent with the approach taken in the NRC-approved Vogtle and Millstone submittals.
Therefore, the NRC staff finds the PFM analysis in the EPRI reports 14590 and 15906 to be appropriate for the ANO-1 and ANO-2 alternative requests.
3.2.3 Parameters Most Significant to PFM Results The NRC staff reviewed the ANO-1 and ANO-2 alternative requests for potential divergence from the Vogtle and Millstone submittals, concerning parameters most significant to PFM results in the EPRI reports. The staff confirmed that the staffs review conclusions in the Vogtle and
Millstone SEs applied to the ANO-1 and ANO-2 alternative requests. The staff found that the parameters most significant to PFM results would be the same and consistent with the staffs reviews documented in the Vogtle and Millstone SEs. Consequently, the approach taken in the Vogtle and Millstone reviews appropriately applies to the review for the ANO-1 and ANO-2 alternative requests.
As discussed in the Vogtle and Millstone SEs, and the sensitivity analysis and sensitivity study in the EPRI Reports, the NRC staff identified the following significant parameters or aspects of the PFM analyses that warrant a close evaluation of the ANO-1 and ANO-2 alternative requests:
stress analysis, fracture toughness, flaw crack growth (FCG) rate coefficient, effect of ISI schedule and examination coverage, and flaw density of weld and nozzle inside radius. The staff evaluated the topics of interest as shown below.
3.2.4 Stress Analysis To evaluate the licensees stress analysis, the NRC staff reviewed the selection of components and materials, the selection of transients, the other operating loads, and the finite element analysis as discussed below.
3.2.4.1 Selection of Components and Materials In attachment 1 of the June 6, 2024, submittal, the licensee evaluated the plant-specific applicability of the components and materials selected and analyzed in the EPRI reports to the subject SG welds and nozzle inside radii at ANO-1 and ANO-2. The acceptability of meeting the criteria, however, depends on the acceptability of the component and material selection described in the EPRI reports, which the NRC staff evaluated below.
Section 4, Survey of Components and Selection of Representative Components for Analysis, of the two EPRI reports discusses the variation among generic SG shell and SG nozzle designs.
EPRI used this information to perform finite element analyses to determine stresses in the analyzed components. 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. EPRI concluded that variations in the configuration of SG lower heads, shells, and nozzles among the various designs are not significant, and that the most important parameter, ratio of radius-to-thickness (R/t) of the primary side nozzles, primary side lower head, secondary side shell, and secondary side nozzles, can be addressed through sensitivity studies on stress in the PFM evaluations in the EPRI reports. The NRC staff notes that R and t are the radius and wall thickness of the subject SG component, respectively.
The NRC staff finds that the SG configurations selected in the EPRI reports for stress analysis are acceptable representatives for the corresponding SG components in the ANO-1 and ANO-2 alternative requests because differences in R/t ratios in the SG configurations discussed in both EPRI Reports are small and, therefore, differences in stresses would be reasonably addressed through the sensitivity study on stress. To verify the dominance of the R/t ratio, the staff reviewed the through-wall stress distributions in section 7, Component Stress Analysis, of the EPRI reports to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. For some of the SG shell welds modeled in EPRI report 15906, the NRC noted that the thermal stress is also potentially high, as discussed in the Millstone SE.
However, the staff determined that EPRI report 15906 is adequate for the subject ANO-1 and ANO-2 SG welds because they are adequately bounded by the sensitivity studies on stress in
EPRI report 15906. Accordingly, the staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for the ANO-1 and ANO-2 alternative requests. The NRC staff determined that the SG configurations at ANO-1 and ANO-2 meet the SG configuration criteria in section 9.4 of the two EPRI reports.
Section 9.4 of the EPRI reports also address criteria for the plant-specific applicability with respect to SG component materials. The plant-specific SG component materials are acceptable if they conform to the ASME Code,Section XI, nonmandatory appendix G, paragraph G-2110.
The licensee addressed these criteria in the submittal. The NRC staff verified that SG materials at ANO-1 and ANO-2 conform with the ASME Code,Section XI, nonmandatory appendix G.
Therefore, the staff finds that the materials for ANO-1 and ANO-2 SGs meet the material applicability criterion. Based on the discussion above, the NRC finds that ANO-1 and ANO-2 SGs are acceptable with respect to the SG component configuration and materials analyzed in the EPRI reports.
3.2.4.2 Selection of Transients Section 5.2, Operating Loads and Transients, of the EPRI Reports discusses the thermal and pressure transients under normal and upset conditions considered relevant to the subject SG components. EPRI developed a list of transients for analysis applicable to the SG components based on transients that have the largest temperature and pressure variations.
The licensee stated that in the selection of the transients in section 5 and the subsequent stress analyses in section 7 of both EPRI Reports, test conditions beyond a system leakage test were not considered since pressure tests at ANO-1 and ANO-2 are performed at normal operating conditions. The licensee further stated that no hydrostatic testing had been performed at ANO-1 and ANO-2 since the units went into operation.
The NRC staff confirmed that the applicable aspects of the transients discussed in the Vogtle and Millstone SEs apply equally to ANO-1 and ANO-2. The staff determined that the transient selection defined in the EPRI Reports is reasonable for the ANO-1 and ANO-2 alternative requests because the selection was based on large temperature and pressure variations that are conducive to fatigue crack growth and are expected to occur in PWRs.
The NRC staff noted that the absence of safety injection sources that inject directly into the primary side of the SG means that the possibility of water that is relatively colder than the reactor coolant at normal operating temperature causing a rapid temperature decrease (i.e.,
thermal shock) of the primary side SG shell is highly unlikely. The NRC staff confirmed that the transients associated with the ANO-1 and ANO-2 SGs are bounded by the transient criteria in the EPRI reports.
Based on the discussion above, the NRC staff finds that the licensee meets the transient applicability criteria in the EPRI reports. Therefore, the analyzed transient loads for the subject SG components of ANO-1 and ANO-2 are acceptable.
3.2.4.3 Other Operating Loads The EPRI reports consider weld residual stress and cladding stresses as part of loading. The NRC staff documented the review of these aspects of the EPRI reports in the Vogtle and Millstone SEs. The staff confirmed that no ANO plant-specific aspects warranted additional consideration, noting that (1) the relatively low sensitivity of the weld residual stress on EPRI
report results as shown in table 8-12 of the EPRI reports; and (2) the small impact of clad residual stress on the PFM results. Based on this, the NRC staff finds that there is a low probability that plant-specific aspects of other operating loads would have a significant effect on the probability of leakage or rupture beyond the studies documented in the EPRI reports.
Based on the discussion above, the NRC staff finds that the treatment of other loads is acceptable for the SG components of ANO-1 and ANO-2.
3.2.4.4 Finite Element Analysis The NRC staff reviewed the finite element analyses in the EPRI reports and documented its review in the Vogtle and Millstone SEs. The staff confirmed that the finite element analyses in the EPRI reports are applicable to the SG components at ANO-1 and ANO-2. Therefore, the staff finds that no plant-specific aspects at ANO-1 and ANO-2 with respect to finite element analyses warranted further review. Based on this finding, the NRC staff determined that the pressure and thermal stresses calculated through finite element analyses in the EPRI reports are acceptable to be applied to the subject SG components of ANO-1 and ANO-2.
3.2.5 Fracture Toughness The EPRI reports assume fracture toughness value 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, appendix A, A-4200. The EPRI reports treat KIC as a random parameter normal distribution with a mean value of 200 ksiin and a standard deviation of 5 ksiin, stating that these assumptions are consistent with the BWRVIP-108NP-A report. The NRC staff evaluated the fracture toughness value used in the EPRI reports in the Vogtle and Millstone SEs. The staff confirmed that the fracture toughness evaluations documented in the Vogtle and Millstone SEs apply to the ANO-1 and ANO-2 alternative requests.
The EPRI reports indicate that the materials of the SG vessel head and tubesheet need to be low alloy ferritic steels, which conform to the requirements of ASME Code,Section XI, appendix G, paragraph G-2110, and which requires information on reference nil-ductility transition temperature (RTNDT). Table 1-1 in attachment 1 to the enclosure of the June 6, 2024, submittal, indicates that the maximum RTNDT value for the ANO-1 SG shell material is 0 degrees Fahrenheit (°F) and is bounded by the value used in the EPRI reports. Table 2-1 in attachment 2 to the enclosure of the June 6, 2024, submittal, indicates that the ANO-2 analysis assumes an RTNDT value of 60oF, which is the maximum allowed by NRC Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 3, July 2018, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Report of Nuclear Power Plants (ML18071A066), and is consistent with the EPRI reports.
The NRC staff determined that the ANO alternative requests meet the material criteria in the EPRI reports, and the EPRI report fracture toughness parameters are applicable to ANO-1 and ANO-2. The staff finds the fracture toughness models in the EPRI reports are acceptable for use for the SG welds and nozzle inside radii at ANO-1 and ANO-2 because the plant-specific RTNDT values are bounded by those used in the EPRI reports.
3.2.6 Flaw Density Section 8.2.2.2 of t EPRI report 14590 and section 8.3.2.2 of EPRI report 15906 assume a flaw density of 0.001 flaw per nozzle for the SG nozzle inside radius. However, the NRC staff has
accepted a flaw density of 0.1 flaw per nozzle at the SG nozzle inside radius as documented in the Vogtle SE because the staff has determined that the increased probabilities of leak and rupture from the lower flaw density assumption are still significantly below the acceptance criterion of 1x10-6 per year. The staff confirmed that the flaw density of 0.1 flaw per nozzle at the SG nozzle inside radius applies to the ANO-1 and ANO-2 alternative requests without further plant-specific evaluations.
The flaw density in the SG welds (1.0 flaw per weld) is based on the flaw density that the NRC staff determined acceptable as documented in the SE for BWRVIP-108 dated December 19, 2007 (ML073600374). Using this flaw density and estimated volumes of the subject SG welds, the staff finds that the assumed flaw density for the subject SG welds is reasonable.
Based on its confirmation that the materials and geometric criteria are acceptable for the subject SG welds and nozzle inside radii at ANO-1 and ANO-2, the NRC staff finds that the appropriate flaw density has been considered, and therefore acceptable, for the subject SG welds and nozzle inner radii of ANO-1 and ANO-2.
3.2.7 Fatigue Crack Growth Rate The NRC staff reviewed the FCG rate used in the EPRI reports and documented its review in the Vogtle and Millstone SEs. The licensee stated that DFM evaluations in table 8-31 of EPRI report 14590 and table 8-3 of EPRI report 15906 provide verification of the above PFM results for ANO-1 and ANO-2 by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code,Section XI, acceptance standards to grow to a depth where the maximum stress intensity factor exceeds the ASME Code,Section XI, allowable fracture toughness.
The NRC staff confirmed that no plant-specific aspects of the ANO-1 and ANO-2 alternative requests warranted further review with regards to FCG rate. Therefore, the staff finds that the ASME Code,Section XI, A-4300 FCG rate used in the EPRI reports is acceptable for the subject SG welds and nozzle inside radii at ANO-1 and ANO-2.
3.2.8 ISI Schedule and Examination Coverage For the ANO-1 replacement SGs installed in 2005, the licensee performed PSI examinations during the third ISI interval, one completed ISI examinations during the fourth interval, and a partial ISI examination during the fifth ISI interval.
For the ANO-2 replacement SGs installed in 2000 (first period of the third inspection interval),
the licensee performed PSI examinations during the third ISI interval, followed by ISI examinations in the third ISI interval, and a partial ISI examination during the fourth ISI interval.
The licensee provided plant-specific ISI results of the subject SG welds and nozzle inside radii in attachments 1 and 2 to the enclosure of the June 6, 2024, submittal. The licensee stated that it did not detect any recordable indications. The inspection history in table 1-5 in attachment 1 to the enclosure of the submittal shows that weld No.03-102 has the lowest examination coverage (73.75 percent) of subject SG welds at ANO-1. Table 2-6 in attachment 2 to the enclosure of the submittal shows that weld No.04-002 has the lowest examination coverage (69.60 percent) of subject SG welds at ANO-2. The NRC staff noted that the sensitivity study in table 8-33 of EPRI report 15906 showed that an examination coverage of 50 percent resulted in a probability of rupture that is less (lower) than the acceptance criterion of 1x10-6 per year. Therefore, the staff
determined that, provided that these two welds met the configuration, materials, and transient selection criteria, an examination coverage of 69.60 and 73.75 percents is acceptable because the probability of rupture associated with these two welds is less than the acceptance criterion of 1x10-6 per year. The NRC staff finds the ISI scenarios of the subject SG welds and nozzle inside radii to be acceptable and adequately represented by the ISI scenarios analyzed in the EPRI reports.
3.2.9 Other Considerations The NRC staff reviewed the initial flaw depth and length distribution, probability of detection, various analytical models, uncertainty in calculations, convergence of simulation, flaw density, and DFM analysis of the EPRI reports and documented its review in the Vogtle and Millstone SEs.
Regarding probability of detection, the licensee stated that both EPRI reports, and the plant-specific evaluations above used an ASME Code,Section XI, appendix VIII-based probability of detection (POD) curve in the PFM evaluation because most ISI examinations of major plant Class 1 and 2 components are performed using appendix VIII procedures. The licensee stated that for Class 2 components, the use of appendix VIII procedures is plant-specific. The licensee further stated that in the case of ANO-1 and ANO-2, ASME Code,Section V, procedures may be used for at least some Class 2 components (e.g., the ANO-1 main steam nozzle inside radius section examinations and the ANO-2 feedwater nozzle inside radius section examinations). Based on the NRCs observations in section 3.8.8.2 of the Vogtle SE, the use of the ASME Code,Section XI, appendix VIII based POD curve for inspections performed under the ASME Code,Section V procedures would have minimal impact of the PFM results because the POD curve is not one of the parameters that significantly affect the PFM results.
The NRC staff confirmed that the plant-specific aspects of these topics do not need further evaluation because these parameters and plant-specific analyses in the ANO-1 and ANO-2 alternative requests are consistent with that of the EPRI reports. Therefore, the NRC staff finds that the ANO-1 and ANO-2 alternative requests are acceptable with regards to these aspects as used in the EPRI reports.
3.2.10 PFM Results Relevant to Proposed Alternatives The PFM analyses in both of the EPRI reports investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. For the proposed alternatives, the licensee performed the following PFM analysis for ANO-1 and ANO-2 with the following assumptions in terms of inspections.
The licensee stated that EPRI report 14590 indicates that the limiting component for Item Nos. C2.21 and C2.22 is the feedwater nozzle. The licensee explained that there are no Item Nos. C2.21 and C2.22 components for the ANO-1 SG feedwater nozzle configuration.
Therefore, the licensee used the main steam nozzles at ANO-1 in lieu of the feedwater nozzle in its PFM analysis. The licensee performed five separate evaluations with the following limiting PSI/ISI scenarios:
(1) ANO-1 main steam nozzle inside radius sections (PSI+10+40+70) [PSI plus one 10-year ISI examinations to be followed by two 30-year ISI deferrals],
(2) ANO-1 main steam nozzle-to-shell welds (PSI+10+40+70),
(3) ANO-2 feedwater nozzle inside radius sections (PSI+10+20+50),
(4) ANO-2 feedwater nozzle to-shell welds (PSI+10+20+50), and (5) the remainder of the ANO-1 and ANO-2 SG welds (PSI+10+40+70).
The licensee stated that the above limiting PSI/ISI scenarios for ANO-1 and ANO-2 were not specifically considered in the EPRI reports PFM evaluations in combination with key variables, as evaluated by the NRCs Vogtle SE. Therefore, the licensee performed additional plant-specific evaluations with the limiting PSI/ISI scenarios shown above.
The licensees PFM evaluation for ANO-1 and ANO-2 demonstrated that the probability of rupture is below the acceptance criterion of 1x10-6 per year. The 80-year PoF results of these additional PFM sensitivity studies are all below the acceptance criterion of 1x10-6 per year. The licensee stated that the PFM results in the EPRI reports indicated that after a PSI followed by subsequent ISIs, no other inspections of the subject SG components are required for the remaining two ISI intervals of the 60-year operating license to meet the PoF acceptance criterion of 1x10-6 per year.
The NRC staff has concerns with the licensees conclusion that the PFM results indicate that the subject SG components do not need to be inspected for the remaining fifth and sixth ISI intervals. The staff finds that the licensees analysis does not account for the effect of the combination of the most significant parameters or the added uncertainty of low probability events. More significantly, the staff considers the licensees conclusion to be a risk-based approach, which is inconsistent with NRC policy that calls for risk insights to be considered together with other factors rather than sole reliance on risk-based approaches. The NRC policy is based on a risk-informed approach, not a risk-based approach. The staff determines that post-fabrication examinations are critical in supporting necessary performance monitoring goals including monitoring and trending, bounding uncertainties, validating/confirming analytical results, and providing timely means to identify novel and/or unexpected degradation. The staff finds that the licensee should include adequate performance monitoring as part of the basis for its proposed alternatives. As such, the NRC staff evaluated the licensees proposed performance monitoring plan below.
3.2.11 Performance Monitoring The NRC staff notes that performance monitoring, such as ISI programs, is a part of the five principles of risk-informed decision-making in RG 1.174. Analyses, such as PFM, work in concert with performance monitoring to provide a mutually supporting and diverse basis for maintaining safe operation. In the context of the proposed alternatives, an adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation/confirmation of continued adequacy of associated analyses, and a timely method to detect novel/unexpected degradation. The staff presented these characteristics at a public meeting on March 4, 2022, with the agenda and presentation slides, which can be found in ADAMS at Accession Nos.ML22053A171 and ML22060A277, respectively.
The NRC staff noted that each unit (i.e., ANO-1 and ANO-2) has two SGs. The footnotes to the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1, Examination Categories B-B, C-A, and C-B permits the volumetric examination to be limited to one vessel among the group of vessels performing a similar function. This means that the licensee is permitted to examine SG components in one SG at ANO-1 and one SG at ANO-2 during each ISI interval. However, Examination Category B-D requires that all welds in all SGs at a plant be examined in an ISI interval. Based on the provisions in tables IWB-2500-1 and IWC-2500-1, ANO-1 is required to
examine 10 SG components and ANO-2 is required to examine 11 SG components during each ISI interval.
The NRC staff requested the licensee to describe how its performance monitoring will provide, over the extended examination interval, (1) direct evidence of the presence and extent of degradation, (2) validation and confirmation of the continued adequacy of the PFM model; and (3) timely detection of novel or unexpected degradation.
In its supplement dated October 16, 2024, the licensee proposed to examine one SG weld at ANO-2 during the sixth ISI interval and to use examinations of SG components at Waterford Steam Electric Station Unit 3 (Waterford) to support the ANO-1 and ANO-2 alternative requests.
By letter dated March 18, 2024 (ML24078A376), the licensee submitted Waterford Alternative WF3-RR-24-02 proposing an alternative to the ASME Code required examinations of SG components. In a supplement dated September 24, 2024 (ML24268A296), the licensee stated that it plans to examine a total of five SG components at Waterford during the fifth and sixth ISI intervals at part of sample population for the performance monitoring plan at ANO-1 and ANO-2.
In its February 20, 2025, supplement, the licensee revised its ANO performance monitoring plan. The licensee proposed to examine two SG welds at ANO-2 during the sixth ISI interval.
The NRC staff evaluated the proposed ANO performance monitoring plan in terms of SG Equivalent. The staffs position is that when the 25 percent of the SG Equivalent (in table 1 below) derived from the ASME Code required SGs examinations is less than the SG Equivalent (in Table 2 below) derived from the licensee proposed SG examinations, then the performance monitoring would be adequate to provide reasonable assurance of structural integrity of the SGs. Table 1 below provides the SG Equivalent derived from the total number of SGs that are required per the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1 to be examined at ANO-1, ANO-2 and Waterford in the fourth, fifth and sixth ISI intervals. Table 2 below provides the SG Equivalent derived from the number of SG components that the licensee proposed to examine relative to the number of SG components that need to be examined per the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1. The licensee also presented the same calculations as shown in its February 20, 2025, supplement.
Table 1: ASME Code Required Number of SGs to be Examined in Fourth, Fifth and Sixth Intervals Site
- of nuclear units
- SGs required to be examined per unit
= units x SGs required per unit x intervals ANO-1 1
1 2
2 ANO-2*1 1
1.375 2
2.75 Waterford*2 1
1 3
3 Total 7.75 Note *1 -- ANO-2 is required to examine eight SG components (welds and nozzle inner radii) in each SG during each ISI interval that are classified under Examination Categories B-B, B-D, C-A and C-B of the ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1. The ASME Code permits welds under Categories B-B, C-A and C-B in one SG to be examined among multiple SGs in the plant during each ISI interval. However, welds under Examination
Category B-D, Item No. B3.130 are required to be examined in all SGs during each ISI interval.
Each ANO-2 SG has three B3.130 welds. As such, ANO-2 needs to examine three out of eight components of the second SG in addition to examining the eight components in the first SG.
This is equivalent to 1.375 SG that the licensee needs to examine in every ISI interval (8/8 + 3/8 = 1.375).
Note *2 -- Waterford examined one SG weld during the fourth ISI interval and has requested to defer SG component examinations for the fifth and sixth ISI intervals. As such, Waterford will be considered as deferring examinations for three ISI intervals (i.e., fourth, fifth and sixth intervals).
Note *3 - ANO-1 and ANO-2 have completed their fourth ISI interval examination. Both plants have requested to defer their SG examinations for the fifth and sixth ISI intervals.
As shown in table 1, above, the licensee is required to examine a total of 7.75 SGs for all three plants. The NRC staffs position on the performance monitoring plan in terms of alternative examination is that licensees are permitted to inspect at a minimum 25 percent of the ASME Code required SGs based on a statistical binominal approach such that there is reasonable assurance that the structural integrity of the SGs in the population will be maintained. This means that the licensee is permitted to examine a total of 1.9375 SG (7.75 SG x 25%) in lieu of 7.75 SG.
Table 2 below shows the number of SG components that the licensee needs to examine to satisfy the 1.9375 SG Equivalent.
Table 2: Number of SG Equivalent Exams for Fifth and Sixth Interval Unit ASME Code required Number of SG Components to be examined Licensee proposed to examine number of SG Components in 5th and 6th intervals SG Equivalents =
Proposed exams
/ required exams ANO-1*1 10 6
0.6 ANO-2*2 11 2
0.18 Waterford*3 5
6 1.2 Total 1.98 Note *1 - ANO-1 had examined six components during the fifth ISI interval.
Note *2 - ANO-2 proposed to examine two welds during the sixth ISI interval.
Note *3 -- Waterford proposed to examine four welds in the fifth interval and one weld in the sixth interval. Waterford inspected one weld in the fourth interval. Therefore, Waterford will have examined a total of 6 components.
Table 2 shows that the licensee will have examined 1.98 SG Equivalent as part of the proposed performance monitoring plan, which is greater than 1.9375 SG Equivalent. Therefore, the NRC staff determined that the licensees proposed performance monitoring plan for the ANO alternative requests is acceptable.
The NRC staff noted that the proposed examination of Waterford SG components is not part of the original ANO-1 and ANO-2 alternative requests. The licensee submitted the Waterford alternative request earlier than the ANO-1 and ANO-2 alternative requests in a separate licensing action. As such, the staff reviews and processes the ANO-1 and ANO-2 and the Waterford alternative requests in separate regulatory actions. Since the proposed SG
examinations at Waterford are not part of the original ANO-1 and ANO-2 alternative requests, the staff requested that licensee provide a performance monitoring plan specifically for the ANO-1 and ANO-2 alternative requests if the Waterford SG examinations do not occur as planned. As shown in table 6 of the February 20, 2025, supplement to the June 6, 2024, submittal, the licensee proposed a contingency plan to examine seven SG components at ANO-2 during the sixth ISI interval in case the SG examinations at Waterford were not performed. The licensee derived the seven SG components to be examined at ANO-2 based on the SG Equivalent calculations. The staff determined that the proposed examination of seven SG components at ANO-2 will satisfy the staffs position on the performance monitoring plan in case the Waterford SG examinations are not available. Therefore, the NRC staff finds that the licensees contingency performance monitoring plan is acceptable.
The NRC staff finds that the licensees proposed performance monitoring plan, including the proposed SG examinations at Waterford, will examine adequate number of SG components to monitor the structural integrity of the SGs and, therefore. are acceptable.
The NRC staff notes that the above finding is only applicable to Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01. The NRC finds that the satisfactory finding on the proposed performance monitoring plan for the ANO-1 and ANO-2 alternative requests is not applicable to the Waterford Alternative Request WF3-RR-24-02, which is being reviewed and processed in a separate regulatory action.
3.2.12 Degradation Management As part of performance monitoring review, the NRC staff also evaluated how the licensee manages potential degradation in the subject SG components that exceeds the acceptance standards of the ASME Code,Section XI, IWB-3500 to ensure that the structural integrity of the subject SG components is adequately maintained. The staff noted that if detected indications at SG components exceed the acceptance standards of the ASME Code,Section XI, IWB-3500, scope expansion may be needed to assess extent of condition. In addition, if a performance monitoring plan or industry-wide operating experience indicates that a new or novel degradation mechanism is possible in SG welds or nozzle inner radii, scope expansion may also be needed to ensure that no such degradation mechanism is occurring in the subject plants.
In its supplement dated October 16, 2024, the licensee stated that if indications are detected that exceed the applicable ASME Code,Section XI, acceptance standards of IWB-3500, it will address the indications as required by the ASME Code,Section XI, and its Corrective Action Program. The licensee further stated that the additional examination and successive inspection requirements of the ASME Code,Section XI, also apply during the current outage. The licensee clarified that the number of additional examinations shall be the number required by the ASME Section XI, IWB-2430. In the February 20, 2025, supplement, the licensee further clarified that it will also follow the provisions of IWC-2430 and IWC-3600.
The licensee stated that it uses its Corrective Action Program to review and evaluate industry operating experience to determine the appropriate actions required. The licensee further stated that if the operating experience indicates that a new or novel degradation mechanism is possible in SG welds or nozzle inner radii, it will perform or consider the appropriate examinations to ensure that no such mechanism is occurring in the subject plants.
In its supplement dated February 20, 2025, the licensee clarified that if a flaw is found as a part of any performance monitoring plan provided in its submittal, it will expand scope to the other
units for the examination category and item number for which the flaw was found as applicable per plant design within the first to second scheduled refueling outages of discovery of the indication. The licensee stated that it will perform scope expansion in accordance with IWB-2430, IWB-3500, IWC-2430 and IWC-3500 as applicable.
The NRC staff has determined that if an unacceptable flaw is detected in the subject SG components, the licensee will inspect SGs in other units in the proposed alternatives within first to second scheduled refueling outages of discovery of the flaw. The NRC staff finds that the proposed scope expansion inspection and associated timeline are consistent with the staff position and is, therefore, acceptable.
In summary, based on PFM and DFM of the EPRI reports and the adequate performance monitoring plan, the NRC staff determined that inspections for the subject SG components at ANO-1 and ANO-2 could be deferred during the proposed period while maintaining an adequate level of plant quality and safety.
4.0 CONCLUSION
Based on information submitted, the NRC staff has determined that the licensees Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01, including supplements dated October 16, 2024, and February 20, 2025, provide an acceptable level of quality and safety. Accordingly, the 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 Alternative Request ANO1-ISI-24-01 at ANO-1 up to May 20, 2034, and Alternative Request ANO2-ISI-24-01 at ANO-2 up to July 17, 2038.
The NRC staff notes that the above conclusion is not applicable to the Waterford Alternative Request WF3-RR-24-02. The staff review and processing the Waterford Alternative Request is a separate regulatory action.
All other ASME Code,Section XI, requirements for which relief has not been specifically requested and approved in the proposed alternative requests remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: John Tsao, NRR Steven Levitus, NRR Date: April 29, 2024
- by email NRR-028 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DNRL/NPHP/BC*
NAME DGalvin PBlechman ABuford DATE 04/15/2025 04/17/2025 04/10/2025 OFFICE NRR/DORL/LPL4/(A)BC NAME TNakanishi DATE 04/29/2025