ML25041A281
| ML25041A281 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/19/2025 |
| From: | Tony Nakanishi NRC/NRR/DORL/LPL4 |
| To: | Entergy Operations |
| Drake J, NRR/DORL/LPL4 | |
| References | |
| EPID L-2024-LLR-0021, WF3-RR-24-01 | |
| Download: ML25041A281 (1) | |
Text
February 19, 2025 Site Vice President Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - PROPOSED ALTERNATIVE WF3-RR-24-01 INSERVICE INSPECTION INTERVAL EXTENSION FOR PRESSURIZER WELDS (EPID L-2024-LLR-0021)
Dear Site Vice President:
By letter dated March 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24078A374), as supplemented by letter dated September 24, 2024 (ML24268A295), Entergy Operations, Inc. (the licensee), submitted a request for Waterford Steam Electric Station, Unit 3 (Waterford 3), to the U.S. Nuclear Regulatory Commission (NRC) for a proposed alternative to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI examination requirements.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed to forgo ASME Code,Section XI examination requirements for the requested pressurizer welds. The regulation in 10 CFR 50.55a(z)(1) requires the licensee to demonstrate that the proposed alternative provides an acceptable level of quality and safety. The NRC staff reviewed the licensees proposed alternative request for Waterford 3 as a plant-specific alternative.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation that the proposed alternative for the requested components provides 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 the proposed alternative, WF3-RR-24-01, at Waterford 3 through the end of the current licensed operating life.
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 me the Waterford 3 Project Manager, Jason Drake at 301-415-8378 or via email at Jason.Drake@nrc.gov.
Sincerely, Tony T. Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No(s). 50-382
Enclosure:
Safety Evaluation cc: Listserv Tony T.
Nakanishi Digitally signed by Tony T. Nakanishi Date: 2025.02.19 16:46:54 -05'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE WF3-RR-24-01 INSERVICE INSPECTION INTERVAL EXTENSION FOR PRESSURIZER WELDS ENTERGY OPERATIONS INC.
WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382
1.0 INTRODUCTION
By letter dated March 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24078A374), as supplemented by letter dated September 24, 2024 (ML24268A295), Entergy Operations, Inc. (Entergy, the licensee), submitted a request for Waterford Steam Electric Station, Unit 3 (Waterford 3) to the U.S. Nuclear Regulatory Commission (NRC) for a proposed alternative to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI examination requirements.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed to forgo ASME Code Section XI examination requirements for the requested pressurizer (PZR) welds. The regulation in 10 CFR 50.55a(z)(1) requires the licensee to demonstrate that the proposed alternative provides an acceptable level of quality and safety.
The NRC staff reviewed the licensees proposed alternative request for Waterford 3 as a plant-specific alternative.
2.0 REGULATORY EVALUATION
The PZR pressure-retaining welds at Waterford 3 are ASME Code Class 1 components, whose inservice inspections (ISIs) are performed in accordance with the applicable edition of Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME Code, as required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements.
The regulations in 10 CFR 50.55a(g)(4), Inservice inspection standards requirements for operating plants, state, in part, that, components that are classified as ASME Code Class 1, 2, and 3 must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state, in part that alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a may be used when authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensees Proposed Alternative
=
Applicable Code Edition and Addenda===
The Code of record for Waterford 3 during the fourth 10-year ISI interval is the 2007 Edition of the ASME Code,Section XI through the 2008 Addenda.
ASME Code Components Affected ASME Code Class:
Class 1 Examination Category:
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 - Pressurizer, Shell-to-Head Welds, Circumferential B2.12 - Pressurizer, Shell-to-Head Welds, Longitudinal B3.110 - Pressurizer, Nozzle-to-Vessel Welds Table 1: Component IDs ASME Category ASME Item No.
Component ID Component Description B-B B2.11 05-002 Bottom head to lower shell circ. Weld B-B B2.11 05-008 Top head to shell weld B-B B2.12 05-003 Lower shell long. weld 90-degrees(1)
B-B B2.12 05-004 Lower shell long. weld 180-degrees(1)
B-B B2.12 05-006 Upper shell long. weld 90-degrees(1)
B-B B2.12 05-007 Upper shell long. weld 180-degrees(1)
B-D B3.110 05-009 Surge nozzle to bottom head weld B-D B3.110 05-010 Spray nozzle to top head weld B-D B3.110 05-011 Safety nozzle to top head weld 45-degrees B-D B3.110 05-012 Safety nozzle to top head weld 135-degrees B-D B3.110 05-013 Safety nozzle to top head weld 225-degrees Note 1 of the table denotes that the applicable portion of the longitudinal seam welds are where they intersect the associated shell-to-head weld. This means that only one out of each pair (lower and upper shell) of welds must be inspected every interval.
ASME Code Requirement for Which Alternative Is Requested For ASME Code Class 1 welds in the PZR, the ISI requirements are those specified in subarticle IWB-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric examinations as specified in ASME Code,Section XI, table IWB-2500-1, for each examination category and item number listed below, and once every 10-year ISI interval.
Examination Category B-B, Item No. B2.11, PZR Shell-to-Head Welds, Circumferential Examination Category B-B, Item No. B2.12, PZR Shell-to-Head Welds, Longitudinal Examination Category B-D, Item No. B3.110, PZR Nozzle-to-Vessel Welds The NRC staff confirmed that the ASME Code requirements listed above did not change in the latest edition of ASME Code,Section XI, incorporated by reference in 10 CFR 50.55a, Codes and standards.
Reason for Proposed Alternative In section 4.0 of the enclosure to the submittal dated March 18, 2024, the licensee stated that the Electric Power Research Institute (EPRI) performed assessments in the following non-proprietary report of the basis for the ASME Code,Section XI examination requirements for the PZR welds identified in this safety evaluation (SE).
EPRI Technical Report 3002015905, Technical Bases for Inspection Requirements for PWR [Pressurized-Water Reactor] Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, Final Report, 2019 (hereinafter referred to as EPRI report 15905 (ML21021A271)).
The 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 licensee stated that this report was developed consistent with EPRIs White Paper on PFM (ML19241A545) and Regulatory Guide (RG) 1.245, Preparing Probabilistic Fracture Mechanics Submittals, Revision 0, January 2022 (ML21334A158). Based on the conclusions of the report, the licensee requested an alternative to the ASME Code,Section XI examination requirements for the subject PZR welds.
The NRC staff noted that EPRI report 15905 was not submitted or reviewed as a topical report.
The staff reviewed the proposed alternative request for the subject plant as a plant-specific alternative. The NRC did not review the EPRI report for generic use, and this review does not extend beyond the plant-specific authorization.
Proposed Alternative and Duration In section 5.0 of the enclosure to the March 18, 2024, submittal, the licensee requested to defer the examinations for Item Nos. B2.11, B2.12, and B3.110 through the end of the current licensed operating life, which is scheduled to end on December 18, 2044. The licensee stated that Waterford 3 has its original pressurizer. According to the licensee, the subject PZR welds and nozzles received the required preservice inspection (PSI) examinations followed by ISI examinations through the first period of the current fourth inspection interval. The licensee further explained that the proposed alternative equates to an extension of 27 years, 18 days from the end of the third ISI interval.
Basis for Proposed Alternative In section 5.1 of the enclosure to the March 18, 2024, submittal, the licensee referred to the results of the PFM analyses in EPRI report 15905 mentioned above and additional PFM sensitivity studies as the bases for the proposed alternative.
3.2
NRC Staff Evaluation
The NRC staffs review focused on evaluating the applicability of the PFM analyses in section 8.3 of EPRI report 15905 and verifying whether the DFM and PFM analyses in the report support the proposed alternative. The staff previously reviewed a similar request based on EPRI report 15905. That request was in support of a Salem Nuclear Generating Station, Unit Nos. 1 and 2 submittal (ML20218A587, hereafter Salem submittal). As part of the previous review of the Salem submittal, the staff conducted a thorough review of the applicable aspects of the EPRI report and documented its review in the associated, plant-specific Salem SE dated June 10, 2021 (ML21145A189). For the Entergy review, the staff considered the referenced information and focused on the plant-specific application of the EPRI report for Waterford 3.
Using a risk-informed approach, the staff also confirmed that the proposed alternative provides sufficient performance monitoring.
3.2.1 Degradation Mechanisms The NRC staff reviewed the licensees submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to Waterford 3 to merit additional consideration. The staff found no evidence of conditions at Waterford 3 that would require consideration of a unique degradation mechanism beyond application of the information the licensee referenced from EPRI report 15905. Specifically, the staff reviewed the materials, stress states, and the consistency of a chemical environment (i.e., reactor coolant) of the subject PZR welds and found them to be consistent with the assumptions made in the EPRI report. Therefore, the staff finds that consideration of additional degradation mechanisms beyond those from the EPRI report is not necessary.
3.2.2 PFM Analysis The NRC staff confirmed that the PFM analysis referenced by the licensee is consistent with the approach taken in the technical arguments presented in the Salem submittal and explicitly referenced in the alternative request. The original review of this approach is documented in the Salem SE. The NRC reviewed the application of this approach, as proposed in the Entergy request, and determined that the PFM analysis is consistent with the previously approved precedents in the Salem submittal. Therefore, the staff finds the proposed PFM analysis to be appropriate for this application for Waterford 3.
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 staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events and other similar reviews. In that rule, the reactor vessel through-wall crack frequency (TWCF) of 1x10-6 per year for a pressurized thermal shock (PTS) 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 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). This assumption is conservative because a through-wall crack in the reactor vessel does not necessarily increase the likelihood of core damage. The discussion of TWCF is explained in detail in the technical basis document for 10 CFR 50.61a, NUREG-1806 Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61),
August 2007 (Package ML072830074).
The NRC staff also noted that the TWCF criterion of 1x10-6 per year was generated using a very conservative model for reactor vessel cracking. In addition, this criterion exists within a context of reactor pressure vessel surveillance programs and inspection programs. The staff finds that the licensees use of 1x10-6 failures per year based on the reactor vessel TWCF criterion is acceptable for the requested PZR welds of Waterford 3 because (a) the impact of a PZR vessel failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject welds have substantive, relevant, and continuing inspection histories and programs; and (c) the estimated risks associated with the individual welds are mostly much lower than the system risk criterion (i.e., the system risk is dominated by a small subpopulation which can be considered the principal system risk for integrity). The staff further noted that comparing the probability of leakage to the same criterion is conservative because leakage is less severe than rupture. The use of a PoF criterion such as 1x10-6 per year for individual welds may not be appropriate generically, but based on the discussion above, the staff finds the application of this criterion acceptable for this plant-specific review for the PZR welds for Waterford, Unit 3.
Lastly, the NRC staff noted that the acceptance criterion of 1x10-6 failures per year is lower, and thus more conservative, than the criterion the staff accepted in proprietary report BWRVIP-05, BWR [Boiling-Water Reactor] Vessel and Internals Project [BWRVIP]: BWR Reactor Pressure Vessel Weld inspection Recommendation, September 1995; 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 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 EPRI reports were developed prior to or around the time the rules for PTS were reevaluated, and as such the acceptance criterion for 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 staff also noted that the BWR Vessel and Internal Project topical reports included substantive inspection aspects that were critical to the NRCs findings.
Based on the discussion above, the NRC staff finds the used of acceptance criterion of 1x10-6 failures per year for PoF acceptable for the Waterford 3 plant-specific alternative request.
3.2.3 Parameters Most Significant to PFM Results The NRC staff reviewed the licensees submittal for plant-specific aspects of the Waterford 3 alternative request that may diverge from the Salem submittal, as explicitly referenced in the Entergy request, concerning parameters most significant to PFM results in EPRI report 15905.
Because Waterford 3 is a Combustion Engineering (CE) design compared to the Westinghouse design used at Salem, the staff reviewed the plant-specific PFM analysis and determined that the parameters most significant to PFM would be the same as those in the EPRI analyses because PZRs were considered in the EPRI analyses.
As discussed in the Salem SE, the sensitivity analysis, sensitivity study, and the NRC staffs observations on the PRobabilistic OptiMization of InSpEction (PROMISE) software identified the following significant parameters or aspects of the PFM analyses that warrant a close evaluation:
stress analysis, fracture toughness, flaw density, flaw crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage. The staff discussed and closely evaluated each in the next five sections of this SE. The staff also evaluated other parameters or aspects of the analyses in section 3.2.9 of this SE.
3.2.4 Stress Analysis 3.2.4.1 Selection of Components and Materials In attachment 1 of the submittal dated March 18, 2024, the licensee evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI report 15905 to the subject PZR welds of Waterford 3. The licensee stated that most plant-specific criteria, as specified in the EPRI report, was met. The acceptability of meeting the criteria, however, depends on the acceptability of the component and material selection described in the EPRI report, which the NRC staff evaluated below.
In section 4 of EPRI report 15905, EPRI discussed the variation among PZR designs. EPRI used this information for finite element analyses (FEA, see section 3.2.4.4 of this SE) to determine stresses in the analyzed components, which the licensee referenced for the corresponding PZR components requested for Waterford 3. 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 15905 and finds that the PZR configurations selected in the report for stress analysis are acceptable representatives for the corresponding PZR components requested for the Waterford 3, plant-specific alternative request. Specifically, the radius-to-thickness (R/t) ratios of the requested Waterford components, provided in tables 1 and 2 of the enclosure to the licensees submittal, are either bounded by the R/t ratios analyzed in the EPRI report or by the sensitivity study on stress in the EPRI report. To verify the dominance of the R/t ratio, the staff reviewed the through-wall stress distributions in section 7 of the EPRI report to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. The staff confirmed that the stresses at Waterford 3 are less than those considered in Westinghouse-designed plants, and thus, the stress analysis in EPRI report 15905 bounds Waterfords CE design. 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 Waterford 3 plant-specific alternative request.
Additionally, in table 1-1 of the March 18, 2024, submittal, the licensee states that the Waterford 3 PZR shell/head and nozzles meet the applicability criteria in EPRI report 15905 regarding weld and nozzle configuration. The NRC staff confirmed that the EPRI report criteria regarding weld and nozzle configuration are met.
Section 9.4 of EPRI report 15905 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. The licensee addressed these criteria in table 1-1 of its submittal. The licensee stated that the PZR vessel heads and shell are fabricated from SA-533 Grade B Class 1, and the nozzles are fabricated from SA-508 Class 2.
The NRC staff verified that these materials conform with ASME Code,Section XI, Nonmandatory Appendix G. Therefore, the staff finds that the materials for the Waterford 3 PZR meet the material applicability criterion.
3.2.4.2 Selection of Transients In section 5.2 of EPRI report 15905, EPRI discussed the thermal and pressure transient under normal and upset conditions considered relevant to the PZR shell and associated welds. EPRI developed a list of transient for analysis applicable to the PZRs analyzed in the report, based on transients that have the largest temperature and pressure variations.
The NRC staff evaluated the transient selection in the EPRI report in detail, as discussed in the Salem SE. As previously stated, Waterford 3 is a CE design compared to the Westinghouse design used at Salem. The staff confirmed that the applicable aspects of the transients discussed in the Salem SE apply equally to this review for Waterford 3, noting that CE-designed PZRs were considered in the EPRI report when selecting general transients, and that a single group of insurge/outsurge transients was developed in the report that applies to both CE and Westinghouse designs. The staff reviewed the discussion of transients in section 5.2 of EPRI report 15905 and determined that the transient selection defined in the report are reasonable for the Waterford 3 plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG and are expected to occur in PWRs. The staff then compared the analysis in the EPRI report to plant-specific information provided in the licensees submittal.
In tables 1-2, 1-3, and 1-4 of the March 18, 2024, submittal, the licensee evaluated the plant-specific applicability of the transients selected in the EPRI report 15905 to the PZR of Waterford 3. The NRC staff reviewed these tables and confirmed that the Waterford PZR shells and nozzles are bounded by the criteria in the EPRI report. The staff noted that there were minor differences in temperatures and pressures. However, the staff determined that these minor variations would not substantially impact the stress calculations underlying the fatigue crack growth calculation. Furthermore, the projected 60-year cycles in tables 1-3 and 1-4 of the submittal are substantially below the number of cycles assumed in the analysis.
In the analyses in the EPRI report, there were no separate test conditions included in the transient selection. The licensee stated on page 6 of 16 of the enclosure to the March 18, 2024, submittal that pressure tests (i.e., system leakage tests) at Waterford 3, are performed at normal operating conditions and no hydrostatic testing has been performed since the plant began operation. The NRC staff noted that since the pressure tests are performed at normal operating conditions, they are part of heatup/cooldown, and therefore test conditions need not be analyzed as a separate transient.
Based on the discussion above, the NRC staff finds that Waterford 3 meets the transient applicability criteria in EPRI report 15905. Therefore, the analyzed transient loads for the requested PZR components of Waterford are acceptable.
3.2.4.3 Other Operating Loads Weld residual stress and cladding stresses are addressed in EPRI report 15905. The NRC staff documented the review of these aspects of the EPRI report in the Salem SE. The staff confirmed that no Waterford 3 plant-specific aspects of this submittal warranted additional consideration, noting (1) the relatively low sensitivity of the EPRI results on residual stress (table 8-14 of EPRI report 15905) and sensitivity studies conducted on stress; and (2) the small impact of clad residual stress on the PFM results. Based on this, the staff finds that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on the probability of leakage or rupture beyond the studies documented in the EPRI report.
Based on the discussion above, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested PZR welds of Waterford 3.
3.2.4.4 Finite Element Analysis The NRC staff reviewed the FEA conducted in EPRI report 15905 and documented its review in detail in the Salem SE. The staff confirmed that no Waterford 3 plant-specific aspects of this application warranted further review. Based on this, the staff determined that the pressure and thermal stresses calculated through FEA in EPRI report 15905 are acceptable for referencing for the requested PZR welds of Waterford 3.
3.2.5 Fracture Toughness In EPRI report 15905, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC value of 200 ksiin based on the upper-shelf fracture toughness value in ASME Code,Section XI, A-4200. EPRI treated 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-108 project. Further discussion of this topic, as it relates to the EPRI report and to plant-specific applications, is contained in the Salem SE. The NRC staff confirmed that the evaluations documented in the Salem SE apply to the Entergy submittal without further plant-specific considerations. As discussed in section 3.2.4 of this SE, Waterford 3, meets the material criteria in EPRI report 15905, and thus the staff has determined that the assumed fracture toughness parameters above are applicable to Waterford 3.
Based on the discussion referenced above and the discussion in section 3.2.4 of this SE, which confirmed that the materials are acceptable for the requested PZR welds of Waterford 3, the NRC staff finds the fracture toughness model in the referenced EPRI report acceptable for the requested PZR welds at Waterford 3.
3.2.6 Flaw Density In section 5.0 of the enclosure to the March 18, 2024, submittal, the licensee stated that a flaw density of 1.0 per weld was used in the plant-specific evaluation consistent with that used in EPRI report 15905. This flaw density is based on the flaw density the NRC staff determined acceptable as documented in the December 19, 2007, SE for BWRVIP-108 (ML073600374).
Using this flaw density and estimated volumes of the subject PZR welds, the staff finds that the assumed flaw density for the PZR welds is reasonable.
Based on the discussion referenced above and the discussion in section 3.2.4 of this SE, which confirmed that the materials and geometric criteria are acceptable for the requested PZR welds of Waterford, Unit 3, the NRC staff finds the appropriate flaw density has been considered, and therefore acceptable, for the requested PZR welds of Waterford 3.
3.2.7 Fatigue Crack Growth Rate The NRC staff reviewed the FCG rate used in EPRI report 15905 and documented its review in detail in the Salem SE. The staff confirmed that no plant-specific aspects of the licensee submittal warranted further review with regards to FCG rate. Based on the discussions referenced above, the staff finds that the ASME Code,Section XI, A-4300 FCG rate used in EPRI report 15905 is acceptable for the requested PZR welds at Waterford 3.
3.2.8 ISI Schedule and Examination Coverage EPRI analyzed various ISI schedules (or scenarios) in chapter 8 of EPRI report 15905. The NRC staff reviewed the applicable aspects of the ISI schedule and examination coverage modeling used in the EPRI report and documented its review in detail in the Salem SE.
The licensee provided information on the inspection history of the requested PZR welds of Waterford 3, in table 1-5, in attachment 1 of the enclosure to the March 18, 2024, submittal, for the third and fourth ISI intervals. The NRC staff determined that the licensees use of only the more recent (i.e., during the third and fourth ISI intervals) examination coverage to be reasonable because the coverages are similar across the two intervals except for the B3.110 welds, which are not the limiting welds in terms of the PFM results. Also, use of the more recent examination coverage is reasonable because for PZR welds, the probability of rupture is relatively insensitive to examination coverage. Table 1-5 indicates that components05-010, 05-011, and 05-012 (all nozzle-to-vessel welds) had recordable indications in at least one inspection. The licensee stated that in every case the indications were acceptable per ASME Code,Section XI, IWB-3600. Inspections for weld 05-010 showed acceptable indications in the inspection performed in the third ASME Code ISI interval, but no indications were observed in the fourth ASME Code ISI interval.
Finally, the inspection history shows that some of the examination coverages did not meet the ASME Code,Section XI examination coverage requirement of 90 percent or greater. However, licensees are required to submit a relief request under 10 CFR 50.55a(g)(5)(iii), ISI program update: Notification of impractical ISI Code requirements, for ASME Code,Section XI examination requirements that are determined by the licensee to be impractical, which typically includes examination coverages that do not meet the requirement. The NRC staff noted an examination coverage as low as 48.3 percent. The staff discussed the impact of this low examination coverage on the PFM results in section 3.2.10 of this SE.
Based on this discussion, the NRC staff finds the Waterford 3, inspection history of the subject PZR welds to be acceptable. Thus, given the discussion above on the inspection history of the requested PZR welds of Waterford 3, the staff finds that the PFM analyses of EPRI report 15905 adequately represents the requested components for Waterford 3, with respect to ISI schedule and examination coverage.
3.2.9 Other Considerations The NRC staff reviewed the application and associated references concerning initial flaw depth and length distribution, probability of detection, models, uncertainty, convergence, flaw density, and DFM analysis. The staff previously reviewed the applicable aspects of these topics as used in EPRI report 15905 and documented its review in detail in the Salem SE. The staff confirmed that no plant-specific aspects of the submittal warranted further review. Based on the discussion referenced above, the staff finds that the licensees submittals are acceptable with regards to these modeling aspects used in the EPRI report, and therefore, is acceptable for the requested PZR welds of Waterford 3.
3.2.10 PFM Results Relevant to Proposed Alternative In section 5.0 of the enclosure to the March 18, 2024, submittal, the licensee stated that the PFM results in EPRI report 15905 indicated that after a PSI followed by subsequent ISIs, the criterion of 1x10-6 failures per year is met. The NRC staff does not find this conclusion acceptable since it 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 this conclusion to be a solely risk-based approach inconsistent with NRC policy that calls for risk insights to be considered together with other factors rather than sole reliance on risk-based approaches. 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 evaluated the licensees proposed performance monitoring plan in section 3.2.11 of this SE.
The licensee evaluated the PFM scenarios on pages 6 through 16 of the enclosure to the March 18, 2024, submittal, to determine which scenario best represent the inspection history of Waterford 3. Because the Waterford 3 PZR is original and has received PSI examinations and examinations for three complete ISI intervals, the PSI/ISI scenario for Waterford is (PSI+10+20+30+60).
This scenario was not specifically considered in EPRI report 15905, so the licensee performed a plant-specific PFM evaluation using PROMISE 2.0, the same version as in EPRI report 15905.
The evaluation was performed for the critical Case ID of PRSHC-BW-2C with a combination of stress and fracture toughness, as identified in the Salem SE. The licensee reported that this analysis resulted in PoFs less than the acceptance criterion of 1x10-6 per year.
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. The NRC staff described these characteristics at various public meetings (ML22060A277, ML23033A667, and ML23114A034).
The initial proposed alternative for Waterford 3, would have resulted in an insufficient number of examinations and a significant amount of time before another examination was performed on all welds under the submittal. The NRC staff requested the licensee to provide additional information regarding a performance monitoring plan that will verify that the assumptions of the PFM analysis remained valid throughout the period of the proposed alternative (see Request for Additional Information (RAI) RAI-1 in the September 24, 2024, supplement).
In response to RAI-1, the licensee described 10 examinations to be performed throughout the period of the proposed alternative. Each of the ASME Code item numbers covered by the proposed alternative (B2.11, B2.12, B3.110) will be examined as part of the proposed performance monitoring plan. The licensees proposed performance monitoring plan schedule is shown in table 1 of the enclosure to the September 24, 2024, supplement, reproduced below as table 2. The ASME Code Section XI requirement is to perform 3 PZR examinations for the alternative period (1 PZR x 3 ISI intervals), with 1 PZR examination consisting of 9 item number examinations (for a total of 27 required item number examinations during the alternative period).
The licensee has proposed to perform 10 of the 27 required item number examinations, which is 37 percent for the proposed alternative, which equates to over one full PZR examination equivalent over the period of the alternative.
Table 2: Proposed Monitoring Plan.
Item No.
Comp ID Exam Date Interval/Period/Outage Future Exam Schedule Outage Current Schedule Approximate Years Between Exams B2.11 05-002 1/12/2019 4th / 1st / RF22 6th Int / 3rd Period 2046 27 B2.11 05-008 10/28/2012 3rd / 2nd / RF18 5th Int / 2nd Period 2032 19.5 B2.12 05-003 1/12/2019 4th / 1st / RF22 6th Int / 3rd Period 2046 27 B2.12 05-007 10/28/2012 3rd / 2nd / RF18 5th Int / 2nd Period 2032 19.5 B3.110 05-009 1/13/2019 4th / 1st / RF22 6th Int / 1st Period 2046 27 B3.110 05-010 1/13/2019 4th / 1st / RF22 6th Int / 1st Period 2046 27 B3.110 05-011 1/11/2019 4th / 1st / RF22 6th Int / 1st Period 2046 27 B3.110 05-012 1/11/2019 4th / 1st / RF22 6th Int / 1st Period 2046 27 B3.110 05-013 1/11/2019 4th / 1st / RF22 6th Int / 1st Period 2039 20 Table 2 above is reproduced as it appeared in the September 24, 2024, supplement, sent by the licensee. The NRC staff notes that the stated future examination period for components05-009, 05-010,05-011, 05-012, and 05-013 is the first period of the sixth ISI interval. However, the year of examination for component 05-013 is 2039 while the other named components is 2046. An examination date of 2046 is given for components05-002 and 05-003 in table 2 as well, which is stated to be during the third period of the sixth ISI interval. The staff also notes that 2046 is after the current licensing period for Waterford 3, and after the proposed end of the subject alternative of December 18, 2044. In making its safety finding, the staff only considered weld examinations that were stated to occur prior to the end of the current licensing life. Thus, any examinations planned for 2046 were not considered in this safety finding.
The NRC staff determined, through binomial statistics and Monte Carlo methods that a 25 percent sample of the total ASME Code required number of PZRs would be an adequate performance monitoring sample over the subject alternative period. In the context of this alternative request this would lead to a sample of 0.25x3 = 0.75 PZRs. However, at least one PZR equivalent of examinations must be performed at minimum over the length of the alternative. As discussed above, the licensee has proposed to do an equivalent of at least one PZR worth of examinations. Consequently, the NRC staff found that the quantity of examinations over the subject alternative period is acceptable.
The NRC staff reviewed the timing of examinations to ensure that the proposed examinations in the performance monitoring plan would provide a reasonable continuous source of data that would support the characteristics of acceptable performance monitoring. Specifically, data would continue to become available on a cadence reasonably commensurate with ASME Code requirements. Based on the proposed examinations during the alternative period, the staff finds that the examinations proposed in the performance monitoring plan will provide an adequate amount of data.
As part of the proposed performance monitoring plan in the supplement dated September 24, 2024, the licensee described actions it would take if degradation was discovered as part of performance monitoring activities. The licensee stated that detected indications would be evaluated and dispositioned according to the rules of ASME Code,Section XI. The licensee stated that domestic and international operating experience would be entered into the Entergy Corrective Action Program to determine appropriate actions.
Based on the above discussion and given the supplemental information in the RAI response, the NRC staff determined that inspections for the subject components could be deferred during the proposed period because an adequate level of performance monitoring is maintained for the components.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the licensees proposed alternative for the requested components provides an acceptable level of quality and safety. Accordingly, the staff concludes that the license 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 WF3-RR-24-01 for Entergy through the end of the current licensed operating life at Waterford 3.
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.
Principal Contributor: C. Parker, NRR D. Dijamco, NRR Date: February 19, 2025
- concurrence via email NRR-028 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*
NRR/DNRL/NVIB/BC*
NRR/DORL/LPL4/BC*
NAME JDrake PBlechman ABuford TNakanishi DATE 2/10/2025 2/12/2025 2/5/2025 2/19/2025