ML26065A290

Relief Request 74 - Proposed Alternative for Steam Generator and Pressurizer Pressure Retaining Welds (EPID L-2025-LLR-0084) - Letter

ML26065A290
Person / Time
Site:	Palo Verde   (NPF-041, NPF-051, NPF-074)
Issue date:	03/18/2026
From:	Michael Mahoney Plant Licensing Branch IV
To:	Heflin A Arizona Public Service Co
Orders, William
References
EPID L-2025-LLR-0084
Download: ML26065A290 (0)

Text

March 18, 2026 Mr. Adam Heflin Executive Vice President/

Chief Nuclear Officer Mail Station 7605 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

Subject:

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 -

RELIEF REQUEST 74 - PROPOSED ALTERNATIVE FOR CERTAIN STEAM GENERATOR AND PRESSURIZER PRESSURE RETAINING WELDS (EPID NO. L-2025-LLR-0084)

Dear Mr. Heflin:

By letter dated September 4, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25247A295), as supplemented by letter dated January 30, 2026 (ML26034C112), Arizona Public Service Company (APS, the licensee) submitted a request for U.S. Nuclear Regulatory Commission (NRC) authorization of proposed inservice inspection (ISI) alternative Relief Request 74 for Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(z)(1).

This alternative defers the required inspections of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Table IWB-2500-1, Examination Categories B-B, B-D, C-A and C-B for Item numbers B2.31, B3.40, B3.130, C1.10, C1.20, C1.30, C2.21 and C2.22 associated with certain steam generator (SG) welds for the remainder of the fourth interval, the fifth and sixth intervals to the end of the current renewed operating licenses scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2 and November 25, 2047, for Unit 3.

This alternative also defers the required inspections of ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-B and B-D for Item numbers B2.11, B2.12, associated with certain pressurizer (PZR) welds for the remainder of the fourth interval, the fifth and sixth intervals to the end of the current operating licenses scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2 and November 25, 2047, for Unit 3.

As set forth in the enclosed safety evaluation, the NRC staff has determined that APS proposed alternative in Relief Request 74 for the requested PZR and SG welds provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that APS has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative for the remainder of the current fourth ISI interval through the end of the sixth ISI interval to the end of the current operating licenses for PVNGS, Units 1, 2 and 3.

A. Heflin All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this alternative request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact the PVNGS project manager William Orders, at (301) 415-3329 or via email at William.Orders@nrc.gov.

Sincerely, Michael Mahoney, Acting Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, 50-529, and 50-530

Enclosure:

Safety Evaluation cc: Listserv MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.03.18 16:29:54 -04'00'

Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ALTERNATIVE REQUEST RR 74 ARIZONA PUBLIC SERVICE COMPANY PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2 AND 3 DOCKET NOS. 50-528, 50-529, AND 50-530 EPID: L-2025-LLR-0084

1.0 INTRODUCTION

By letter dated September 4, 2025 (Agencywide Documents Access and Management System Accession No. ML25247A295), as supplemented by letter dated January 30, 2026 (ML26034C112), Arizona Public Service Company (APS, the licensee) submitted a request for U.S. Nuclear Regulatory Commission (NRC) authorization of proposed inservice inspection (ISI) alternative Relief Request 74 for Palo Verde Nuclear Generating Station (Palo Verde, PVNGS), Units 1, 2, and 3, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(z)(1).

This proposed alternative would defer the required inspections of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Table IWB-2500-1, Examination Categories B-B, B-D, C-A and C-B for Item numbers B2.31, B3.40, B3.130, C1.10, C1.20, C1.30, C2.21 and C2.22 associated with certain steam generator (SG) welds for the remainder of the fourth interval, the fifth and sixth intervals to the end of the current renewed operating licenses scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2, and November 25, 2047, for Unit 3.

This proposed alternative would also defer the required inspections of ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-B and B-D for Item numbers B2.11, B2.12, associated with pressurizer (PZR) welds for the remainder of the fourth interval, the fifth and sixth intervals to the end of the current operating licenses scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2, and November 25, 2047, for Unit 3.

The regulation in 10 CFR 50.55a(z)(1) requires APS to demonstrate that the proposed alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

The NRC regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state, in part, that alternatives to the requirements of paragraphs (b) through (h) of this section

[50.55a] or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or NRC report, NUREG-1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) screening limit in the PTS Rule (10 CFR 50.61), dated August 2007 (package, ML072830074), summarizes the results of a 5-year study conducted by the NRC to develop the technical basis for revision of the PTS Rule, as set forth in 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, and is used as the basis for the NRC staff's review, consistent with the NRCs current guidelines on risk-informed regulation.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for APS to request the alternative and the NRC staff to authorize it.

3.0 TECHNICAL EVALUATION

3.1 Relief Request 74 Applicable ASME Code Edition and Addenda The fourth 10-year ISI interval Code of record for Palo Verde is the 2013 Edition of ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

Palo Verde, Unit 1, is currently in its fourth 10-year ISI interval, which began June 1, 2019, and ends July 17, 2028. In addition, Palo Verde, Unit 2, is currently in its fourth 10-year ISI interval, which began November 1, 2019, and ends October 31, 2028. Finally, Palo Verde, Unit 3, is currently in its fourth 10-year ISI interval, which began June 1, 2018, and ends January 10, 2028.

ASME Code Components Affected Steam Generators

==

Description:==

SG pressure-retaining welds and full penetration welded nozzles (nozzle-to-shell welds and inside radius sections)

ASME Code Class: Class 1 and 2 Examination Categories:

Category B-B, pressure-retaining welds in vessels other than reactor vessels Category B-D, full penetration welded nozzles in vessels Category C-A, pressure-retaining welds in pressure vessels Category C-B, pressure-retaining nozzle welds in pressure vessels Item Numbers:

B2.31 - SG (primary side), circumferential head weld B2.40 - SG (primary side), tubesheet-to-head weld B3.130 - SG (primary side), nozzle-to-vessel welds C1.10 - Shell circumferential welds C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell welds C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections Component identifications (IDs): The three tables provided on Pages 1 to 3 in the enclosure of the licensees submittal lists the component IDs affected in the replacement SGs for each unit at PVNGS. Further discussion related to the timing of when the steam generators were replaced in each unit is provided below in safety evaluation (SE) Section 3.2.8.

Pressurizer

==

Description:==

PZR vessel head, shell-to-head, and nozzle-to-vessel welds ASME Code Class: Class 1 Examination Categories:

Category B-B, pressure-retaining welds in vessels other than reactor vessels Category B-D, full penetration welded nozzles in vessels Item Numbers:

B2.11 - PZR, shell-to-head welds, circumferential B2.12 - PZR, shell-to-head welds, longitudinal B3.110 - PZR, nozzle-to-vessel welds Component IDs: The three tables provided in Pages 4 to 5 in the enclosure of the licensees submittal lists the component IDs affected in the PZR for each unit at PVNGS.

Applicable ASME Code Requirements For ASME Code Class 1 welds in the SG, the ISI requirements are those specified in Subarticle IWB-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric examinations as specified in Table IWB-2500-1, for the Examination Category and Item Numbers listed below once every ISI interval. As noted in Table IWB-2500-1 for Examination Category B-B, for cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel among the group of vessels performing a similar function. However, welds in all vessels classified under Examination Category B-D, item No. B3.130 are required to be examined once every ISI interval.

Examination Category B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels o Item No. B2.31 - SG (primary side), circumferential head weld o Item No. B2.40 - SG (primary side), tubesheet-to-head weld Examination Category B-D, Full Penetration Welded Nozzles in Vessels o Item No. B3.130 - SG (primary side), nozzle-to-vessel welds For ASME Code Class 2 welds in the SG, the ISI requirements are those specified in Subarticle IWC-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric and surface examinations as specified in Table IWC-2500-1, for each Examination Category and Item No. listed below once every ISI interval. As noted in Table IWC-2500-1 for Examination Categories C-A and 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.

Examination Category C-A, Pressure Retaining Welds in Pressure Vessels o C1.10 - Shell circumferential welds o C1.20 - Head circumferential welds o C1.30 - Tubesheet-to-shell weld Examination Category C-B, Pressure Retaining Nozzle Welds in Pressure Vessels o C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) weld o C2.22 - Nozzle inside radius section Reason for Proposed Request In its submittal dated September 4, 2025, APS stated that the Electric Power Research Institute (EPRI) performed assessments in the following non-proprietary reports of the basis for the ASME Code,Section XI, examination requirements for the pressurizer and SG welds identified in this SE. Unless otherwise noted hereinafter these reports will be referred to as the EPRI Reports.

Report No. 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, EPRI, Palo Alto, CA: 2019. (ML19347B107)

Report No. 3002015906, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2019.

(ML20225A141).

Report No. 3002015905, Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell to-Head, and Nozzle-to-Vessel Welds, EPRI, Palo Alto, CA:

2019. 3002015905 (ML21021A271)

Based on the conclusions of these EPRI reports supplemented by a plant-specific stress analysis, and plant-specific deterministic fracture mechanics (DFM) and probabilistic fracture mechanics (PFM) evaluations for the Palo Verde Combustion Engineering (CE) System 80 SG design, APS is requesting an ISI examination deferral for the subject welds.

The NRC staff notes that the EPRI Reports were not submitted or reviewed as a topical report.

The NRC staff reviewed the proposed alternative request as a plant-specific alternative. The NRC did not review the EPRI Reports for generic use, and this review does not extend beyond the plant-specific authorization for PVNGS, Units 1, 2 and 3.

Proposed Alternative and Duration APS requested to apply the proposed alternative to defer the ISI examinations for these Item Nos. for the SG and PZR welds at PVNGS from the current ASME Code,Section XI, Division 1, 10-year requirement to for the remainder of the fourth interval, the fifth and sixth intervals to the end of current renewed operating licenses, scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2, and November 25, 2047, for Unit 3.

Basis for Proposed Alternative In its submittal, the licensee discussed the key aspects of the technical basis in the EPRI Reports, as supplemented by plant-specific analyses, and plant-specific applicability of these EPRI Reports to PVNGS. The EPRI Reports were used as basis for the proposed alternative for the ASME Code,Section XI, Examination Categories B-B and B-D pressurizer welds. In addition, the EPRI Reports, as supplemented by a plant-specific stress analysis, and plant-specific DFM and PFM evaluations for the Palo Verde CE System 80 SG design, were used as basis for the proposed alternative for the ASME Code,Section XI, Examination Categories B-B, B-D, C-A and C-B SG welds. The licensee also identified prior NRC approvals of similar alternatives at other facilities that relied on the same or comparable technical basis.

3.2

NRC Staff Evaluation

The NRC staff focused on evaluating the applicability of the PFM analyses in Section 8.3 of the EPRI Report 3002015905 and verifying whether the DFM and PFM analyses in the report support the proposed alternative for the pressurizer welds. APS cited an NRC-approved precedent for its request that was based on EPRI Report 3002015905; specifically, the approval of a precedent for Salem Generating Station (Salem), Units 1 and 2 (ML20218A587). The licensee referenced applicable portions of the technical basis from these submittals. The NRC staff documented its review of these applications in the associated plant-specific SE (ML21145A189). Additionally, the NRC staffs review also focused on evaluating the applicability of the PFM analyses in Section 8 of the EPRI Report 3002014590 and 3002015906, as supplemented by a plant-specific stress analysis, and plant-specific DFM and PFM evaluations for the Palo Verde CE System 80 SG design, and determining whether this information supports the proposed alternative for the steam generator welds. The NRC staff considered the information referenced and focused on the plant-specific application of the EPRI Reports for PVNGS. Consistent with the key principles of the NRC risk-informed approach for performing reviews, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.

3.2.1 Degradation Mechanisms In its submittal dated September 5, 2025, APS stated, in part, that:

An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in the EPRI Reports.

The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG and pressurizer welds covered in this requestThe fatigue-related mechanisms were considered in the PFM and DFM evaluations in the EPRI Reports.

The NRC staff reviewed the APS submittal, as supplemented by letter dated January 30, 2026, for plant-specific circumstances that may indicate presence of a degradation mechanism and circumstances sufficiently unique to PVNGS, to warrant additional consideration. Such circumstances pertain to materials of the subject SG and PZR welds, stress states, and reactor coolant environments. The NRC staff found that the degradation mechanisms described by the licensee for PVNGS, are addressed in a manner sufficient for the applicability of the EPRI Reports, as supplemented by a plant-specific stress analysis, and plant-specific DFM and PFM evaluations for the Palo Verde CE System 80 SG design, and that no unknown degradation mechanisms were identified for subject SG and PZR welds.

3.2.2 PFM Analysis The NRC staff noted that the acceptance criterion of 1x10-6 failures per year (also termed Probability of Failure (PoF)) is tied to that used by the NRC staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events, and other similar reviews. In that rule, the reactor vessel through wall crack frequency (TWCF) of 1x10-6 events per year for a pressurized thermal shock event is an acceptable criterion, because reactor vessel TWCF is conservatively assumed to be equivalent to an increase in core damage frequency, and as such meets the guidance in NRC Regulatory Guide (RG) 1.174, An Approach to for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis. This assumption is conservative because a through-wall crack in the reactor vessel does not necessarily increase the likelihood of core damage. The discussion of TWCF is explained in detail in the technical basis document for 10 CFR 50.61a, NUREG-1806 Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61).. In addition, this criterion exists within the context of reactor pressure vessel surveillance programs and inspection programs.

Pressurizer In the Enclosure to its submittal APS stated, in part, that:

Finite element analysis (FEA) was performed in Reference [9.3] to determine the stresses in the pressurizer welds covered in this request. The analysis was performed using representative Westinghouse plant geometries (which bound CE plants), 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 in the EPRI Report to Palo Verde is demonstrated in Attachments 2 and 3 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the EPRI Report stress analysis are compared to those of Palo Verde.

The licensee also stated, in part, that:

Flaw tolerance evaluations were performed in the EPRI Report consisting of PFM evaluations and confirmatory DFM evaluations. Since the Westinghouse pressurizer configuration considered in the EPRI Report is bounding relative to the CE pressurizer design, the results of the flaw tolerance evaluation can be conservatively applied to the Palo Verde pressurizer. The results of the PFM analyses indicate that, after PSI followed by subsequent ISI, the NRC safety goal of 1.0x10-6 failures per year is met.

In the enclosure to its submittal, APS cited several precedents, including a similar plant-specific request made by PSEG Nuclear to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos. B2.11 and B2.12) and Category B-D (Item No. B3.110) volumetric examinations (ML20218A587), which the NRC authorized in its SE (ML21145A189). The NRC staff confirmed that the plant-specific analysis provided by APS for the PVNGS, is consistent with the approach taken in the PSEG Nuclear submittal for Examination Categories B-B (Item Nos. B2.11 and B2.12) and B-D (Item No. B3.110).

The NRC staff finds that licensees use of 1x10-6 failures per year based on the reactor vessel TWCF criterion is acceptable for its alternative request because (a) the impact of a pressurizer weld failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject PZR welds have established inspection histories and ongoing inspection programs that provide continuing monitoring of degradation; and (c) the estimated risks associated with the individual welds are generally much lower than the system risk criterion (i.e., the overall system risk is driven by a limited subset of welds that represent the dominant contributors to integrity risk).

Based on the discussion above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for APS plant-specific alternative request for the PZR welds.

Steam Generator The staff noted that plant-specific DFM and PFM evaluations were performed for the Palo Verde CE System 80 SG design and are contained in Attachment 4 of the submittal. The licensee explained that the results of the DFM evaluation for the CE System 80 SG configuration are summarized in Table 5 of Attachment 4, which shows that the period required for hypothetical postulated flaws to leak is in excess of 200 years, which is indicative that the CE System 80 SG components are flaw tolerant. The maximum stress intensity factors (K) from the DFM evaluation are also shown in Table 5 of this attachment and show that they are all below the fracture toughness of 220 ksiin (kilopound per square inch square root inch).

The NRC staff finds that licensees use of 1x10-6 failures per year based on the reactor vessel TWCF criterion is acceptable for the licensees alternative request because (a) the impact of a steam generator weld failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject SG welds have established inspection histories and ongoing inspection programs that provide continuing monitoring of degradation; and (c) the estimated risks associated with the individual welds are generally much lower than the system risk criterion (i.e.,

the overall system risk is driven by a limited subset of welds that represent the dominant contributors to integrity risk).

Based on the discussion above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for APS plant-specific alternative request for SG welds.

3.2.3 Parameters Most Significant to PFM Results The NRC staff reviewed the parameters or aspects most significant to the PFM analysis: stress analysis, fracture toughness, flaw density, fatigue crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage. The NRC staff also reviewed the plant-specific applicability of the PFM analyses presented in the EPRI Reports for the pressurizer and the plant-specific PFM analyses for the Palo Verde CE System 80 SG design to PVNGS, as discussed in the sections below.

3.2.4 Stress Analysis 3.2.4.1 Selection of Components and Materials Pressurizer In Attachment 1 of the submittal, APS evaluated the plant-specific applicability of the components and materials selected and analyzed in the EPRI Report 15905 to the subject pressurizer welds. The EPRI Report evaluated representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The report also defined plant-specific applicability criteria related to component geometries, materials, and loading conditions that must be evaluated and met by each plant to determine-the applicability of the report.

The licensee stated that the plant-specific applicability of these requirements were met and that the results and conclusions of the EPRI Report are applicable to PVNGS. The acceptability of meeting these criteria, however, depends on the acceptability of the component and material selection described in the EPRI Report, which the NRC staff evaluated below with respect to PVNGS units. The NRC staff independently evaluated the loading conditions (i.e., transient selection) criteria further in this SE.

In Section 4 of EPRI Report 3002015905, EPRI discussed the variation among PZR designs.

EPRI used this information for finite element analyses to determine stresses in the analyzed components, which the licensee referenced for the corresponding PZR components requested for PVNGS. In selecting the components, EPRI considered geometry, operating characteristics, materials, field experience with respect to service-induced cracking, and the availability and quality of component-specific information.

The NRC staff reviewed Section 4 of EPRI Report 3002015905 for applicability to the PVNGS PZR components cited in the licensees submittal and evaluated whether the configurations analyzed in the EPRI report are representative of the corresponding PVNGS PZR components.

The staff finds that the PZR configurations selected for stress analysis are acceptable representatives for use in this plant-specific alternative request for PVNGS because the radius-to-thickness (R/t) ratios of the requested PVNGS components, provided in Tables 1 and 2 of the enclosures to the submittal, are bounded by the R/t ratios analyzed in the EPRI report. The NRC staff noted that Table 1 is the comparison of R/t ratios for the PZR shell/head welds (Item Nos. B2.11 and B2.12), and Table 2 is the comparison of R/t ratios for the PZR nozzle-to-vessel welds (Item No. B3.110). To verify the dominance of the R/t ratio, the NRC staff reviewed the through-wall stress distributions in Section 7 of the EPRI report and confirmed that the pressure stress is dominant, supporting the conclusion that R/t ratio is the appropriate parameter for comparison. Accordingly, the NRC staff finds that the EPRI evaluation of the R/t ratio as the dominant parameter in evaluating the various configurations is appropriate and acceptable for this plant-specific alternative request.

The licensee cited Section 9 of the EPRI Report 3002015905, which addresses criteria for plant-specific applicability of the analysis and indicates that materials are acceptable if they conform to ASME Code,Section XI, Nonmandatory Appendix G, paragraph G-2110. APS addressed these criteria in Tables 1-1 of Attachment 1 to the submittal for PVNGS. The materials of construction for the PZR components are in Table B below.

Table B: Materials of Construction PZR Component Material top head and shell SA-533 Grade B Class 1 top head and surge nozzles SA-541 Class 2 material The NRC staff verified that these materials of construction reported by the licensee conform with the material property requirements of ASME Code Section XI, Nonmandatory Appendix G, based on a review of the material specifications and information presented in PVNGS final safety analysis report (FSAR), Section 5.2. Additionally, the staff noted that SA-541 Class 2 material is a carbon and alloy steel forging used for pressure vessel components with a specified minimum yield strength of 50 kilopounds per square inch (ksi). The staff determined that the specified minimum yield strength is consistent with the dataset used to develop the ASME Code,Section XI, Nonmandatory Appendix G, fracture toughness curve, and requirements of the ASME Code,Section III, Subsection NB2331, have been met per the FSAR. As such, the NRC staff finds that SA-541 Class 2 material is consistent with the technical basis in the EPRI Report 3002015905. Therefore, the NRC staff finds that the licensees PZR materials meet the material applicability criteria.

Table 1-1 of Attachment 1 to the submittal states that the PZR surge nozzle, bottom head and shell, top head nozzles, and head and shell meet the applicability criteria in the EPRI Report 3002015905 regarding weld configuration. APS provided supplementary drawings in Figures 1-1 through 1-6 of Attachment 1 to the submittal to illustrate the weld geometries. The NRC staff reviewed the licensees information against the applicability criteria and finds that the subject PZRs meet the weld geometry applicability criteria described in Section 4 of the EPRI Report 3002015905, as part of its plant-specific evaluation of the licensees proposed alternative.

Based on the above, the NRC staff finds that PVNGS meet the component geometry and materials applicability criteria in the EPRI Report 3002015905. The analyzed geometries and materials in the EPRI Report 3002015905. are acceptable for the requested PZR welds at PVNGS.

Steam Generator Table 2 and Section 5.1 of Calculation No. 2300243.301 contained in Attachment 3 to the submittal summarizes the materials used in creating the SG model, due to the CE System 80 design, including the material properties information that was obtained from the relevant portions of the 1989 Edition of ASME Code,Section III, Appendices. Based on its review, the staff confirmed that the materials summarized in Calculation No. 2300243.301 that were used in creating the SG model are consistent with the design information presented in FSAR Section 5.2. The NRC staffs review regarding the appropriateness of the fracture toughness curve used by the licensee in its plant-specific DFM and PFM evaluations for the PVNGS SG design is documented below in SE Section 3.2.5 3.2.4.2 Selection of Transients Pressurizer In Attachment 1 to its submittal, APS evaluated the applicability of the transients selected and analyzed in EPRI Report 3002015905 to the subject pressurizer welds. The licensee stated that the plant-specific applicability criteria regarding transients were met. The acceptability of meeting the criteria depends on whether the transient selection described in EPRI Report 3002015905 are appropriate as applied to PVNGS, which the NRC staff evaluated below.

In Section 5.2 of EPRI Report 3002015905, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to PZRs. EPRI developed a list of transients for analysis applicable to the PZRs analyzed, based on transients that have the largest temperature and pressure variations.

The NRC staff evaluated the transient selection in Section 5.2 of the EPRI report 3002015905 as applied to PVNGS and confirmed that the applicable aspects of the transients for PVNGS, Units 1, 2 and 3, are sufficiently representative. The NRC staff determined that the transient selection defined in Section 5.2 of EPRI Report 3002015905 is reasonable for the PVNGS, Units 1, 2 and 3, plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG that are expected to occur in PWRs.

In Tables 1-3 and 1-4 of Attachment 1 to the submittal, APS evaluated the plant-specific applicability of the transients selected in the EPRI Report 3002015905 to the PVNGS pressurizer welds. The NRC staff reviewed these tables and confirmed that the transient projections are bounded by the criteria in the EPRI Report. Furthermore, the projected 60-year cycles in Tables 1-3 and 1-4 of Attachment 1 are below the number of cycles assumed in the EPRI Report. Therefore, the NRC staff finds that the transients assumed in the EPRI Report appropriately bound PVNGS.

In the analyses in the EPRI Report 3002015905, there were no separate test conditions included in the transient selection. APS stated in the Enclosure to its submittal that test conditions beyond a system leakage test were not considered because pressure tests at PVNGS are performed at normal operating conditions and no hydrostatic testing had been performed at PVNGS since the unit went into operation. The NRC staff noted that since the pressure tests are performed at normal operating conditions, it is part of Heatup/Cooldown and therefore test conditions need not be analyzed as a separate transient.

Based on the discussion above, the NRC staff finds that the transient selection, as applied to PVNGS, Units 1, 2 and 3, is technically appropriate. This determination is based on the staffs evaluation of the underlying technical bases for transient identification and bounding conditions described in the EPRI Report 3002015905 and their applicability to PVNGS. Accordingly, the NRC finds that analyzed transient loads for the subject PZR welds are acceptable and that the analyses adequately represent plant conditions at PVNGS.

Steam Generator In Attachments 3 and 4 to its submittal, APS evaluated the following transients listed below, which are consistent with those analyzed in EPRI Reports 3002014590 and 3002015906.

RCS Plant Heatup/Cooldown Pressurizer Heatup/Cooldown Plant Loading at 5%/min Plant Unloading at 5%/min Reactor Trip, Loss of Flow, Loss of Load (Enveloped)

The applicability of these transients to PVNGS depends on the technical adequacy of the transient selection methodology described in EPRI Reports 3002014590 and 3002015906 and its relevance to the PVNGS design and operating conditions. The NRC staff evaluated the underlying bases for transient identification and bounding assumptions in those reports, as applied to PVNGS, discussed below.

In Section 5.2 of EPRI Reports 3002014590 and 3002015906, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to SG shell, and SG nozzles. EPRI developed a list of transients for analysis applicable to the PZRs, SG shell, and SG nozzles based on transients that have the largest temperature and pressure variations. The NRC staff evaluated the transient selection in these two EPRI reports and confirmed that the applicable aspects of the transients for PVNGS SGs are addressed sufficiently, as supplemented by a plant-specific stress analysis, and plant-specific DFM and PFM evaluations for the PVNGS SG design. The NRC staff reviewed the transients in Section 5.2 of EPRI Reports 3002014590 and 3002015906, as supplemented by plant-specific stress analysis, and plant-specific DFM and PFM evaluations. The NRC staff determined that the transient selection defined in the EPRI reports is reasonable and appropriate for the PVNGS plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG that are expected to occur in PWRs and are specific to the PVNGS SG design.

In Table 4 of Attachment 4 to the submittal, APS evaluated the expected number of occurrences for these transients used in its plant-specific DFM and PFM evaluations for the PVNGS SGs.

The NRC staff reviewed Table 4 and confirmed that the transient projections are bounded by the number of cycles used in the plant-specific DFM and PFM evaluations. Furthermore, in some cases the projected 60-year cycles in Table 4 for the PVNGS SGs are substantially below the number of cycles assumed in the EPRI analysis. Therefore, the NRC staff finds that the transients assumed in its plant-specific stress analysis, and plant-specific DFM and PFM evaluations bound the design of the PVNGS SGs.

APS stated on page 11 of the enclosure to its submittal that test conditions beyond a system leakage test were not considered since pressure tests at PVNGS are performed at normal operating conditions and that no hydrostatic testing had been performed at PVNGS since the unit went into operation. The NRC staff noted that since the pressure tests are performed at normal operating conditions, it is part of Heatup/Cooldown. Therefore, test conditions need not be analyzed as a separate transient in APS plant-specific stress analysis, and plant-specific DFM and PFM evaluations.

Based on the discussion above, the NRC staff finds that the analyzed transient loads are acceptable for PVNGS SG welds. The NRC further finds that the modeled transients in the plant-specific analyses are appropriate to the PVNGS SGs and therefore are acceptable.

3.2.4.3 Residual Stresses The NRC staff reviewed the proposed alternative with regards to weld residual stress (WRS) and clad residual stress (CRS). WRS and CRS are generically addressed in the EPRI Reports 3002014590, 3002015905, and 3002015906. The staff evaluated how APS applied the methodologies described in those reports to the PVNGS SGs in its plant-specific stress analysis, and plant-specific DFM and PFM evaluations. With regard to WRS, pressure vessel welds typically receive post-weld heat treatment (PWHT) to reduce the effects of weld residual stresses. In EPRIs evaluation, weld residual stresses remaining after PWHT were characterized in the form of a cosine distribution with a peak stress of 8 ksi. The staff confirmed that this approach is consistent with the approach used in these EPRI Reports and appropriately reflects the effects of PWHT for the subject welds. With regard to CRS, cladding at the affected locations of the steam generator was modeled in the plant-specific stress analyses in of the submittal. Thus, the NRC staff finds the effect of cladding stresses were appropriately addressed for the PVNGS SG design in APS plant-specific stress analysis.

The NRC staff determined that no PVNGS plant-specific aspects of residual stress warranted consideration because (1) APS demonstrated, using the methodologies described in the EPRI Reports and supplemented by a plant-specific stress analysis for the PVNGS SG design, the relatively low sensitivity of the EPRI results on stress (see Table 8-14 of EPRI Report 3002015905, Table 8-12 of EPRI Report 3002015906, and Table 8-12 of EPRI Report 3002014590); and (2) the small impact of clad residual stress on the PFM results in the EPRI Reports, as supplemented by plant-specific DFM/PFM analyses for the PVNGS SG design. The NRC staff finds that there is a very low probability that plant-specific aspects of residual stress would have a significant effect on the probability of leakage or rupture beyond the studies documented in the EPRI Reports, as supplemented by a plant-specific stress analysis and plant-specific DFM/PFM analyses for the PVNGS SG design. Based on the above, the NRC staff finds the treatment of residual stresses in the submittal acceptable for the subject PVNGS pressurizer and steam generator welds.

3.2.4.4 Finite Element Analysis Pressurizer APS performed a finite element analysis (FEA) as discussed in EPRI Report 3002015905 to determine the stresses in the pressurizer welds covered in its alternative request by using representative Westinghouse plant geometries (which bound CE plants), bounding transients, and typical material properties. For the proposed alternative, the NRC staff reviewed material and component selection, transient selection, and assumed residual stresses to determine whether the FEA of the EPRI Report 3002015905 is applicable to the subject welds. The NRC staff determined that APS demonstrated that the FEA in the EPRI Report is applicable to the subject welds in each of these respects. Based on the above, the NRC staff finds that the generic FEA performed in the EPRI Report 3002015905 adequately represents the plant-specific PVNGS pressurizer welds.

Steam Generator In Attachment 3 to its submittal, APS provided a plant-specific finite element model that was used to determine the stresses due to thermal transients with appropriate pressure loading so that through-wall stresses can be extracted at locations of interest to be used in a separate fracture mechanics evaluation. The NRC staff noted that APS needed to develop a plant-specific finite element model because the generic SG configuration used in the stress analyses of EPRI Reports 3002014590 and 3002015906 were for standard Westinghouse/CE SG designs; however, these analyses are not appropriate for the PVNGS SG design. APS explained that the technical approach used in Attachment 3 to its submittal is consistent with the EPRI Reports but used the plant-specific geometry and operating conditions to account for the PVNGS SG design. For the proposed alternative, the NRC staff reviewed material and component selection, transient selection, and assumed residual stresses to determine whether the plant-specific finite element model, including the approach used in the EPRI Reports, is applicable to the subject welds. The NRC staff determined that APS demonstrated that the plant-specific finite element model is applicable to and addresses the subject SG welds in each of these respects. Based on the above, the NRC staff finds that the plant-specific FEA adequately represents the PVNGS SG design subject SG welds.

3.2.5 Fracture Toughness In Attachment 1 to its submittal, APS stated that the materials of the subject PVNGS pressurizer components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.

Additionally, in Attachments 3 and 4 to its submittal, APS provided details for the materials of the subject PVNGS SG components and indicated that these materials in the subject components are all ferritic steels and the fracture toughness curve provided in ASME Code,Section XI, Appendix A, Figure A-4200-1, was used in its evaluation. During its review, the NRC staff confirmed that the materials of the subject PVNGS SG components are specifically addressed in Appendices A and G of the 2019 edition of ASME Code,Section XI, which has been incorporated by reference in 10 CFR 50.55a.

The NRC staff independently verified in the licensees FSAR and the respective material specifications that these materials in the SGs and PZRs conformed to the requirements of ASME Code,Section XI, Paragraph G-2110. In the EPRI Reports 3002014590, 3002015905, and 3002015906, and the plant-specific DFM and PFM evaluations for the PVNGS SG, it is assumed for fracture toughness of ferritic materials an upper-shelf fracture toughness (KIc) value of 200 ksiin based on the KIc curve in the ASME Code,Section XI, A-4200. The A-4200 fracture toughness curve refers to the same fracture toughness curve in the ASME Code,Section XI, Paragraph G-2110. The NRC staff determined that the submittal is acceptable with regards to fracture toughness of subject components because the materials of the subject PZR and SG welds conform to the requirements of ASME Code,Section XI, Paragraph G-2110.

3.2.6 Flaw Density Pressurizer In the enclosure to its submittal, APS described a plant-specific sensitivity study assuming 1.0 flaws per weld in the PZR. The licensee stated that the probabilities of leak and rupture were still significantly below the acceptance criterion of 1x10-6 failures per year. The NRC staff noted that Section 8.3.2.2 of the EPRI Report 3002015905 stated that a flaw density of 1.0 flaws per weld was used in the PZR analysis. The NRC staff noted that, so long as the component materials applicability criteria are met, the use of a flaw density of 1.0 flaws per weld is sufficient for the analysis of the subject PZR welds. APS provided plant-specific information regarding the geometries and materials of the subject PZR welds which met the applicability criteria. Based on the above, the NRC staff finds that the flaw density assumed in the EPRI analysis is appropriate for the subject PZR welds.

Steam Generators In the plant-specific PFM and DFM evaluations that address the PVNGS SG design, the licensee stated that 1.0 flaws per weld was assumed in the steam generator. The NRC staff noted that the use of 1.0 flaws per weld is consistent with the NRC staff position in its SE for BWRVIP-108 (ML19297F806), except at the main steam and feedwater nozzle inner radius (NIR), which used 0.1 flaw per nozzle in the PFM analyses in BWRVIP-108. The NRC staff observed in the audit summary report for PROMISE (ML20258A002) that the flaw density value in the main steam and feedwater NIR is a multiplier on the probability values reported in the software output. Thus, the NRC staff finds the licensees use of 1.0 flaws per weld in the feedwater NIR is conservative in comparison to the NRC staff position in its SE for BWRVIP-108.

3.2.7 Fatigue Crack Growth Rate The NRC staff reviewed the application with regards to FCG rate and noted that the FCG rate used in the EPRI Report 3002015905 and the licensees plant-specific DFM and PFM evaluations in Attachment 4 of the submittal is based on the ASME Code,Section XI, A-4300, FCG rate. The NRC staff noted that FCG rate depends on component material and environmental conditions. Per the ASME Code,Section XI (2021 Edition), the A-4300 FCG rate may be used for low alloy ferritic steels in air and reactor water environments. The licensee provided plant-specific information regarding the materials of the subject SG and PZR components. Since the identified materials meet the scope of A-4300, the NRC staff determined that the A-4300 FCG rate is appropriate to be used in the evaluations of subject SG and PZR components.

3.2.8 Examination History The licensee provided information on the examination history of the subject PZR and SG welds in Tables 2-1 and 2-2 for PVNGS, Unit 1, respectively, Tables 2-3 and 2-4 for PVNGS, Unit 2, respectively and Tables 2-5 and 2-6 for PVNGS, Unit 3, respectively, in the enclosure to the submittal. These tables indicate that there were no unacceptable indications found during these examinations. The licensee explained that acceptable indications noted in the tables were fabrication-related indications, which were found to be code acceptable, and that the ongoing ASME Section XI examinations have not detected age-related growth of the indications.

The licensee stated in its submittal that the PVNGS still has the original PZRs; however, the SGs were replaced on the following schedule:

• Unit 1 - 2005 (third period of the second inspection interval)

• Unit 2 - 2003 (second period of the second inspection interval)

• Unit 3 - 2007 (third period of the second inspection interval)

For the SGs, the licensee described a plant-specific PFM analyzing the limiting PSI/ISI scenario of PSI+10+40+70, where a PSI was performed, an ISI was performed during the first interval after the SG replacement (+10), a proposed ISI 30 years from the first ISI (+40) and followed by another proposed ISI 30 years from the previous ISI (+70). This inspection scenario captures and is representative of the examination history of the PVNGS, replacement SGs.

For the PZR, the licensee described a plant-specific PFM analyzing the limiting PSI/ISI scenario of PSI+10+20+30+60, where a PSI was performed, an ISI was performed during the first interval (+10), a second ISI was performed (+20), a third ISI was performed (+30), and a proposed ISI 30 years from the third ISI (+60). This inspection scenario captures the examination history of the PVNGS, original PZRs, and is considered bounding because it did not take credit for the partial-interval examinations performed during the 4th ISI interval which are not credited.

Based on its review, the NRC staff finds that the licensee sufficiently accounted for plant-specific examination history with the supplementary PFM analyses for the replacement SGs and PZRs.

The NRC staff noted that the examination coverages did not meet the ASME Code,Section XI, examination coverage requirement of 90 percent or greater. However, licensees are required to submit a relief request under 10 CFR 50.55a(g)(5)(iii) indicating impracticality to meet ASME Code,Section XI examination coverage.

The NRC staff noted an examination coverage of at least 96.5% of the required weld volume for all SG components, as indicated in letter dated January 30, 2026, and of at least 65% of the required weld volume for all PZR components, as indicated in letter dated September 4, 2025, which are discussed separately below.

Steam Generators In the SE for Millstone Power Station (Millstone), Unit No. 2 (ML21167A355), the NRC staff documented its evaluation of examination coverage as low as 50 percent. In this SE, the NRC staff determined that the PFM results were not sensitive to examination coverage and determined that the probability of rupture in both scenarios remained below the acceptance criterion of 1x10-6 per year. The subject SG welds at PVNGS have all received more coverage than the limiting case reviewed previously by the NRC staff for Millstone; thus, the NRC staff finds that the PVNGS review is bounded by the staffs previous review and evaluation documented in the Millstone SE. Therefore, the NRC staff finds that the licensees plant-specific examination history for the subject SG welds at PVNGS, with at least 80 percent examination coverage, is sufficiently bounded by the PFM analyses in the EPRI Report 3002015906 and the plant-specific PFM described in the licensees submittal.

Pressurizers In the SE for Salem (ML21145A189), the NRC staff documented its evaluation of examination coverage as low as 37.2 percent. In this SE, the NRC staff determined that the PFM results were not sensitive to examination coverage and determined that the probability of rupture remained below the acceptance criterion of 1x10-6 per year. The NRC staffs technical basis for that conclusion is documented in the Salem SE. The subject PZR welds at PVNGS have all received more coverage than the limiting case reviewed previously by the NRC staff for Salem; thus, the staff finds that the situation in its present review with PVNGS is bounded by the staffs previous review and evaluation documented in the Salem SE. Therefore, the NRC staff finds that the licensees plant-specific examination history for the subject PZR welds at PVNGS, with at least 65% examination coverage, is sufficiently bounded by the PFM analyses in the EPRI Report 3002015905 and the plant-specific PFM described in the licensees submittal.

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 the EPRI Report 3002015905 for the PZR and in the licensees plant-specific DFM and PFM evaluations for the PVNGS SGs, do not depend on plant-specific information. Unlike these considerations mentioned above, parameters such as component geometries, materials, and transient selection are dependent on plant -specific information, and the licensee provided plant-specific information to ensure applicability of the analyses in the EPRI Report 3002015905 for the PZRs, which is discussed in the relevant portions of this SE.

Initial depth and length distribution of analyzed flaw in the weld 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 PZR and SG welds in different plants are subject to the same volumetric examination requirements of the ASME Code,Section XI. The models (e.g., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in fracture mechanics analyses as discussed in the NRC staffs safety evaluation for Millstone, Unit 2.

Uncertainty and convergence do not depend on plant-specific information because these are part of the overall PFM analyses that were addressed in the sensitivity studies and sensitivity analyses in the EPRI Reports 3002015905, 3002014590 and 3002015906 and were also discussed in the licensees plant-specific DFM and PFM evaluations for the PVNGS SGs.

Since these considerations are not dependent on plant-specific information, the NRC staff finds that the plant-specific PVNGS submittal is acceptable in terms of these considerations.

3.2.10 PFM Results Relevant to Proposed Alternative The PFM analysis in the EPRI Report 3002015905 investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. Additionally, the licensee performed a plant-specific PFM analyzing the limiting PSI/ISI scenario for PVNGS SG.

For the PZR welds, the PFM results relevant to the proposed alternative are those that most closely represent the proposed alternative ISI schedules. As described above, APS performed a plant-specific PFM study for PSI+10+40+70 for the SGs, and PSI+10+20+30+60 for the pressurizers, to investigate the proposed examination schedules. The plant-specific PFM results for the SG and the relevant PFM results for the pressurizer show that the probability of rupture is below the acceptance criterion of 1x10-6 failures per year. Based on the above, the NRC staff finds that APS analyses adequately bound the proposed examination schedules.

3.2.11 Performance Monitoring Performance monitoring, such as ISI programs, is a necessary component described by the NRC five principles of risk-informed decision making. Analyses, such as PFM, work along with performance monitoring to provide a mutually supporting and diverse basis for facility condition and maintenance that is within its licensing basis. An adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation of continued appropriateness of associated analyses, and a timely method to detect novel/unexpected degradation.

These characteristics regarding performance monitoring were presented, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277; meeting notice and NRC slides, respectively). The NRC staff has previously applied binomial statistics and Monte Carlo methods to augment evaluation of inspection periods beyond 20 years as well. The methods used by the NRC staff were presented at a May 25, 2022, public meeting (ML22144A345, and ML22143A840, meeting notice and NRC slides, respectively).

Pressurizer The licensee explained that PVNGS collectively completed 33 percent of the required ASME Section XI examinations in the fourth interval to-date and no additional pressurizer examinations will be performed during the current fourth inspection interval for any of the PVNGS units.

The proposed performance monitoring plan for the pressurizers, applicable to the fifth and sixth ISI intervals, is that the licensee will examine PZR welds in one out of the three PVNGS units to complete the required ASME Code,Section XI examinations during each of the fifth and sixth Interval. The licensee clarified that the PVNGS unit selected will distribute the examinations across the inspection periods as required by the ASME Code,Section XI, Table IWB-2411-1 to ensure continuous collection of data and allow for timely identification of any service-induced degradation or the emergence of any novel degradation mechanisms.

The NRC staff noted that the proposed examination schedule will provide performance data on the subject welds throughout the licensed life of the PZR in three units and is conservative relative to the PFM PSI+10+20+30+60 scenario cited by the licensee as its technical basis. The NRC independent review determined that the technical basis for the proposed examination schedule is consistent with past NRC-approved alternative requests as documented in the Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Unit Nos. 1 and 2 (Byron), SE (ML24194A022). Therefore, the NRC staff finds that the proposed alternative provides sufficient performance monitoring for the subject pressurizer welds.

Steam Generators The licensee explained that PVNGS, have collectively completed 51 percent of the required ASME Code Section XI examinations in the fourth interval to-date and no additional SG examinations will be performed during the current fourth inspection interval for any of the PVNGS units. Additionally, the proposed performance monitoring plan for the SGs, applicable to the fifth and sixth ISI intervals, is that the licensee will inspect the welds required by the ASME Code in a single SG in one out of the three PVNGS units to complete the required ASME Code,Section XI examinations during each of the fifth and sixth Intervals. The licensee will also examine two SG tubesheet blowdown nozzles in the other units although they are not the subject of this alternative request. The licensee clarified that the PVNGS unit selected for examination will distribute the examinations across the inspection periods as required by the ASME Code,Section XI, Table IWB-2411-1 or Table IWC-2411-1, as applicable, to ensure continuous collection of data and allow for timely identification of any service-induced degradation or the emergence of any novel degradation mechanisms.

The NRC staff noted that the proposed examination schedule provides performance data on the subject SG welds throughout the licensed life of the SGs in three units and is conservative relative to the PFM PSI+10+40+70 scenario cited by the licensee as its technical basis. The NRC independent review determined that the technical basis for the proposed examination schedule is consistent with past NRC-approved alternative requests, as documented in ML24179A326 (for Braidwood, Byron, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and R. E. Ginna Nuclear Power Plant). Therefore, the NRC staff finds that the proposed alternative provides sufficient performance monitoring for the subject SG welds.

Scope Expansion for Pressurizer and Steam Generators The licensee discusses its plans for future inspections in accordance with the ASME Section XI code of record. The NRC staff notes that scope expansion applies if the licensee discovers unexpected degradation during a performance monitoring examination. The scope expansion provides a method for the licensee to investigate extent of condition, should unexpected degradation occur.

The licensee described actions they would take if new unacceptable indications are identified as part of performance monitoring activities. The licensee stated that detected indications would be evaluated according to the rules of the ASME Code,Section XI (which include additional examination and successive inspection requirements), and the PVNGS corrective action program. These additional activities are described in detail in the enclosure to the submittal.

In addition, the licensee explained that all other inspection activities, including the system leakage test (under Examination Categories B-P and C-H) will continue to be performed in accordance with the ASME Code,Section XI requirements, providing further assurance of safety. The NRC staff notes 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 components in the steam generator or pressurizer.

However, the NRC staff notes that system leakage tests are complementary to performance monitoring provided by ISI examinations, which offer timely identification of through-wall leakage should it occur. While not a substitute for volumetric examinations, these leakage tests compliment the ISI program by providing additional assurance and increases confidence that the proposed examinations will provide an acceptable level of performance monitoring for the subject steam generator and pressurizer components. The NRC staff finds that the licensees proposed future inspections and the scope expansion plans, in concert with the increased level of performance monitoring, are acceptable.

4.0 CONCLUSION

As set forth above, the NRC staff determined that APS proposed alternative in Relief Request 74 for the requested PZR and SG components provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that APS has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative for the remainder of the fourth interval, the fifth and sixth intervals to the end of the current renewed operating licenses scheduled for June 1, 2045, for Unit 1, April 24, 2046, for Unit 2, and November 25, 2047, for Unit 3.

All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this alternative request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: On Yee, NRR John Tsao, NRR Date: March 18, 2026

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, 50-529, AND 50-530 RENEWED OPERATING LICENSE NO. NPF-41 RELIEF REQUEST 74 - PROPOSED ALTERNATIVE FOR STEAM GENERATOR AND PRESSURIZER PRESSURE RETAINING WELDS DATED MARCH 18, 2026 DISTRIBUTION:

PUBLIC RidsACRS_MailCTR Resource RidsNrrDorlLpl4 Resource RidsNrrPMPaloVerde Resource RidsNrrLAPBlechman Resource RidsRgn4MailCenter Resource ADAMS Accession Nos.:

ML26065A292: Package ML26065A290: Letter e-Concurrence case: 20260306-90001