ML25041A302
| ML25041A302 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/20/2025 |
| From: | Tony Nakanishi Plant Licensing Branch IV |
| To: | Entergy Operations |
| Drake, Jason | |
| References | |
| EPID L-2024-LLR-0020 | |
| Download: ML25041A302 (16) | |
Text
February 20, 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-24-02 INSERVICE INSPECTION INTERVAL EXTENSION FOR STEAM GENERATOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES (EPID L-2024-LLR-0020)
Dear Site Vice President:
By letter dated March 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24078A376), as supplemented by letter dated September 24, 2024 (ML24268A296), 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 steam generator welds and nozzles. 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 NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1).
Therefore, the NRC staff authorizes the use of the proposed alternative, WF3-RR-24-02, 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. 50-382
Enclosure:
Safety Evaluation cc: Listserv Tony T.
Nakanishi Digitally signed by Tony T. Nakanishi Date: 2025.02.20 11:24:43 -05'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE WF3-RR-24-02 INSERVICE INSPECTION INTERVAL EXTENSION FOR STEAM GENERATOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES 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. ML24078A376), as supplemented by letter dated September 24, 2024 (ML24268A296), 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 steam generator (SG) welds and nozzles. 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 SG pressure-retaining welds and SG full penetration welded nozzles at Waterford 3, are ASME Code Class 1 and 2 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
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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 and 2 Examination Category:
Category B-B, Pressure-Retaining Welds in Vessels Other than Reactor Vessels Category C-A, Pressure-Retaining Welds in Pressure Vessels Category C-B, Pressure-Retaining Nozzle Welds in Pressure Vessels Item Numbers:
B2.40 - Steam Generators (Primary Side), Tubesheet-To-Head Weld C1.20 - Head Circumferential Welds C1.30 - Tubesheet-to-Shell Weld C2.21 - Nozzle-to-Shell (Nozzle-to-Head or Nozzle-to-Nozzle)
Weld C2.22 - Nozzle Inside Radius (NIR) Sections Table 1: Component IDs ASME Category ASME Item No.
Component ID Component Description B-B B2.40 03-076 Steam Generator (Primary Side)
Tubesheet-to-Head Weld B-B B2.40 04-076 Steam Generator (Primary Side)
Tubesheet-to-Head Weld C-A C1.20 03-068 Pressure Vessels Head Circumferential Welds C-A C1.20 04-068 Pressure Vessels Head Circumferential Welds C-A C1.30 03-075 Pressure Vessels Tubesheet-to-Shell Weld C-A C1.30 04-075 Pressure Vessels Tubesheet-to-Shell Weld C-B C2.21 03-073 Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle-to-Shell (Nozzle-to-Head or Nozzle-to-Nozzle) Weld C-B C2.21 04-073 Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle-to-Shell (Nozzle-to-Head or Nozzle-to-Nozzle) Weld C-B C2.22 03-074 Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle Inside Radius Section C-B C2.22 04-074 Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle Inside Radius Section ASME Code Requirement for Which Alternative Is Requested 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 each examination category and item number listed below and once every 10-year ISI interval. As noted in table IWB-2500-1 for Examination Category B-B, in cases of multiple vessels of similar design, size, and service (such as SGs),
the required examinations may be limited to one vessel or distributed among the vessels.
Examination Category B-B, Item No. B2.40, SG Primary Side Tubesheet-to-Head Welds For ASME Code Class 2 welds and NIR sections 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 number listed below once every 10-year ISI interval. As noted in table IWC-2500-1 for Examination Categories C-A and C-B, in cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.
Examination Category C-A, Item No. C1.20, Head Circumferential Welds Examination Category C-A, Item No. C1.30, Tubesheet-to-Shell Welds Examination Category C-A, Item No. C2.21, Nozzle-to-Shell Welds Examination Category C-A, Item No. C2.22, NIR Sections 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 submittal dated March 18, 2024, the licensee 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 SG welds and nozzles identified in this safety evaluation (SE).
EPRI Technical Report 3002015906, Technical Bases for Inspection Requirements for PWR [Pressurized-Water Reactor] Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 2019 (hereinafter referred to as EPRI report 15906, ML20225A141)).
EPRI Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, 2019 (hereinafter referred to as EPRI report 14590, ML19347B107)).
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 these reports were 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 two reports, the licensee requested an alternative to the ASME Code,Section XI examination requirements for the subject welds.
The NRC staff noted that the EPRI reports were not submitted or reviewed by the NRC as topical reports. The staff reviewed the proposed alternative request for the subject plant 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.
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. B2.40, C1.20, C1.30, C2.21, and C2.22 through the end of the current licensed operating life, which is scheduled to end on December 18, 2044. The licensee stated that the Waterford 3 SGs were replaced in 2012. According to the licensee, the subject SG 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 leads to a maximum period 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 the two EPRI reports mentioned above and additional PFM sensitivity studies as the bases for the proposed alternative. EPRI report 15906 was used as a basis for the proposed alternative for the SG ASME Code Examination Categories B-B and C-A welds. EPRI report 14590 was used as a basis for the proposed alternative for the SG ASME Code Examination Category C-B welds and NIR.
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 15906 and section 8.2 of EPRI report 14590 and verifying whether the DFM and PFM analyses in the reports support the proposed alternative. The staff previously reviewed similar requests based on EPRI reports 15906 and 14590. These requests were in support of a Millstone Power Station Unit 2 submittal (ML20198M682, hereafter Millstone submittal) and a Vogtle Electric Generating Plant, Units 1 and 2, submittal (ML19347B105, hereafter Vogtle submittal). As part of the previous reviews of these submittals, the staff conducted a thorough review of the applicable aspects of the EPRI reports and documented its review in the associated, plant-specific SEs (Millstone (ML21167A355) and Vogtle (ML20352A155)). For the Entergy review, the staff considered the referenced information and focused on the plant-specific application of the EPRI reports for Waterford 3, and determined that the technical bases for decisions in the Millstone and Vogtle SEs still apply to the Waterford 3 submittal. 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 reports 14590 and 15906. Specifically, the staff reviewed the materials, stress states, and the consistency of a chemical environment (i.e.,
reactor coolant) of the subject SG welds and NIR and found them to be consistent with the assumptions made in the EPRI reports. Therefore, the staff finds that consideration of additional degradation mechanisms beyond those from the EPRI reports is not necessary.
3.2.2 PFM Analysis The NRC staff confirmed that the PFM analysis referenced by the licensee is consistent with the approach taken in the technical arguments presented in the Millstone and Vogtle submittals and explicitly referenced in the alternative request. The original review of this approach is documented in the Millstone and Vogtle SEs. The reviewed the application of this approach, as proposed in the Entergy alternative request, and determined that the PFM analysis is consistent with the previously approved precedents in the Millstone and Vogtle submittals. 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 SG welds and NIR of Waterford 3 because (a) the impact of a SG vessel failure would be less than the impact of a reactor vessel failure on overall risk; (b) the subject 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 SG welds and NIR for Waterford 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 BWRVIP 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 Millstone and Vogtle submittals, as explicitly referenced in the Entergy request, concerning parameters most significant to PFM results. The staff confirmed that the review conclusions in the Millstone and Vogtle SEs applied to the Entergy submittal and found that the parameters most significant to PFM results would be the same and consistent with the staffs reviews documented in the Millstone and Vogtle SEs.
Consequently the approach taken in the Millstone and Vogtle reviews appropriately applies to the current review for Waterford 3.
As discussed in the Millstone and Vogtle SEs, the sensitivity analysis, the 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 reports 14590 and 15906 to the subject SG welds and NIR of Waterford 3. The licensee stated that most plant-specific criteria as specified in EPRI reports 14590 and 15906 were met. The acceptability of meeting the criteria, however, depends on the acceptability of the component and material selection described in the EPRI reports, which the NRC staff evaluated below.
In section 4 of EPRI reports 14590 and 15906, EPRI discussed the variation among SG shell and SG nozzle 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 SG 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 reports 14590 and 15906 and finds that the SG configurations selected in the report for stress analysis are acceptable representatives for the corresponding SG components requested for the Waterford 3, plant-specific alternative request.
Specifically, the radius-to-thickness (R/t) ratios of the requested Waterford 3 components, provided in tables 1 and 2 of the enclosure to the licensees March 18, 2024, submittal, are either bounded by the R/t ratios analyzed in the EPRI reports or by the sensitivity study on stress in the EPRI reports. To verify the dominance of the R/t ratio, the staff reviewed the through-wall stress distributions in section 7 of EPRI reports14590 and 15906 to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. For some of the SG shell welds modeled in EPRI report 15906, the staff noted that the thermal stress is also potentially high, as discussed in the Millstone SE. However, because thickness is the controlling parameter for thermal stress (the lower the thickness, the lower the thermal stress), the staff determined that EPRI report 15906 would still be adequate for the corresponding welds for the subject plants because the thickness values of the Waterford 3 SG vessels, as shown in table 1 of the enclosure to the submittal, would be adequately bounded by the sensitivity study on stress in the report. 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 plant-specific alternative request.
Additionally, in table 1-1 of the submittal, the licensee states that the Waterford 3 SG shell and nozzles meet the applicability criteria in EPRI reports 14590 and 15906 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 reports 14590 and 15906 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 March 18, 2024, submittal. The licensee stated that the SG vessel heads, vessel shells, tubesheet, and feedwater nozzles are all fabricated from SA-508 Grade 3 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 Waterford 3 SG meet the material applicability criterion.
3.2.4.2 Selection of Transients In section 5.2 of EPRI reports 14590 and 15906, EPRI discussed the thermal and pressure transient under normal and upset conditions considered relevant to the SG shell and SG nozzles. EPRI developed a list of transient for analysis applicable to the SG shell and SG nozzles analyzed in the report, based on transients that have the largest temperature and pressure variations.
The NRC staff evaluated the transient selection in the EPRI reports in detail, as discussed in the Millstone and Vogtle SEs. The staff confirmed that the applicable aspects of the transients discussed in those SEs apply equally to this review for Waterford 3. The staff reviewed the discussion of transients in section 5.2 of EPRI reports 14590 and 15906 and determined that the transient selection defined in the reports 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 reports to plant-specific information provided in the licensees submittal.
In tables 1-2, 1-3, 1-4, and 1-5 of the March 18, 2024, submittal, the licensee evaluated the plant-specific applicability of the transients selected in the EPRI reports 14590 and 15906 to the SG shell and SG nozzles of Waterford 3. The NRC staff reviewed these tables and confirmed that the Waterford 3 SG shells and nozzles are bounded by the criteria in the EPRI reports. 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 1-2 through 1-5 of the submittal are substantially below the number of cycles assumed in the analysis.
In the analyses in EPRI reports 14590 and 15906, there were no separate test conditions included in the transient selection. The licensee stated on page 7 of 18 of the enclosure to the March 18, 2024, submittal, that pressure tests (i.e., system leakage tests) for 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 the EPRI reports14590 and 15906. Therefore, the analyzed transient loads for the requested SG components at Waterford are acceptable.
3.2.4.3 Other Operating Loads Weld residual stress and cladding stresses are addressed in EPRI reports 14590 and 15906.
The NRC staff documented the review of these aspects of the EPRI reports in the Millstone and Vogtle SEs. 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 8-12 of EPRI report 15906 and table 8-12 of EPRI report 14590) 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 reports.
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 SG welds and NIR of Waterford 3.
3.2.4.4 Finite Element Analysis The NRC staff reviewed the FEA conducted in EPRI reports 14590 and 15906 and documented its review in detail in the Millstone and Vogtle SEs. 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 the EPRI reports 14590 and 15906 are acceptable for referencing for the requested SG welds and NIR of Waterford 3.
3.2.5 Fracture Toughness In EPRI reports 14590 and 15906, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC value of 200 ksiin based on the upper-shelf fracture toughness value in the ASME Code,Section XI, A-4200. 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 reports, and to plant-specific applications, is contained in the Millstone and Vogtle SEs. The NRC staff confirmed that the evaluations documented in the Millstone and Vogtle SEs 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 reports 14590 and 15906, 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 SG welds and NIR of Waterford 3, the NRC staff finds the fracture toughness model in the referenced EPRI reports acceptable for the requested SG welds and NIR of Waterford 3.
3.2.6 Flaw Density In section 8.2.2.2 of EPRI report 14590, EPRI stated that 0.001 flaw per nozzle is assumed at the NIR. The NRC staff noted in the Vogtle SE that the acceptable number of flaws in the NIR is 0.1 flaw per nozzle. Further discussion of this topic as it relates to EPRI report 14590 is contained in the Vogtle SE. The staff confirmed that the evaluation documented in the Vogtle SE applies to the Waterford 3 submittal regarding the SG nozzle welds and NIR included in the request without further plant-specific considerations. As discussed in section 3.2.4 of this SE, Waterford 3, meets the material and geometric criteria in EPRI report 14590, and thus the NRC staff determined that the NIR flaw density parameters are applicable to Waterford. The flaw density in the SG welds is based on the flaw density the 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 SG welds, the staff finds that the assumed flaw density for the SG 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 SG welds and NIR of Waterford 3, the NRC staff finds the appropriate flaw density has been considered, and therefore acceptable, for the requested SG welds and NIR of Waterford.
3.2.7 Fatigue Crack Growth Rate The NRC staff reviewed the FCG rate used in EPRI reports 14590 and 15906 and documented its review in detail in the Millstone and Vogtle SEs. The staff confirmed that no plant-specific aspects of the licensees 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 reports 14590 and 15906 is acceptable for the requested SG welds and NIT of Waterford 3.
3.2.8 ISI Schedule and Examination Coverage EPRI analyzed various ISI schedules in chapter 8 of EPRI reports 14590 and 15906. The NRC staff reviewed the applicable aspects of the ISI schedule and examination coverage modeling used in the EPRI reports and documented its review in detail in the Millstone and Vogtle SEs.
The licensee provided information on the inspection history of the requested SG welds and NIR of Waterford 3, in table 1-6 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 of the high coverages obtained for Waterford 3 SG welds and NIR. This table indicates that there were no unacceptable indications found during these examinations.
The licensee stated in section 5.0 of the enclosure to the licensee submittal that the SGs at Waterford had been replaced. Given the implementation of ISI in the PFM analyses in the EPRI reports, as the staff explained in the Millstone and Vogtle SEs, the staff noted that in terms of PFM modeling, ISIs with replacement would be at least as beneficial as only ISIs because replacement is essentially repair of a postulated flaw, while the outcomes of ISI are either repair of a postulated flaw or non-detection and growth of a postulated flaw.
Based on this discussion, the NRC staff finds the Waterford 3, inspection history of the subject SG welds and NIR to be acceptable. Thus, given the discussion above on the inspection history of the requested SG welds and NIR of Waterford 3, the staff finds that the PFM analyses of EPRI reports 14590 and 15906 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 reports 14590 and 15906 and documented their review in detail in the Millstone and Vogtle SEs. 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 reports, and therefore, is acceptable for the requested SG welds and NIR of Waterford.
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 reports 14590 and 15906 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 7 through 12 of the enclosure to the March 18, 2024, submittal to determine which scenarios best represent the inspection history of Waterford 3. This evaluation depends upon whether the plants SGs were replaced, the date of the PSI examinations, and the dates of subsequent ISI examinations. The licensee did three separate evaluations with the following limiting PSI/ISI scenarios:
The licensee identified that the limiting component for items C2.21 and C2.22 was the FW nozzle. The licensee then performed three sensitivity study evaluations with these limiting scenarios since they were not specifically considered in the EPRI reports using the following cases from the reports: Case ID FEW-P1N for the inner radius of the FW nozzle, Case ID FEW-P3A for the FW nozzle-to-shell welds, and Case ID SGPTH-4A for the SG shell welds. The licensee reported that these analysis cases 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 and NIR 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 five examinations to be performed throughout the period of the proposed alternative. Each of the ASME Code item numbers covered by the proposed alternative (B2.40, C1.20, C1.30, C2.22, C2.21) 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 SG examinations for the alternative period (1 SG x 3 ISI intervals), with 1 SG examination consisting of 5 item number examinations (for a total of 15 required item number examinations during the alternative period). The licensee has proposed to perform 6 of the 15 required item number examinations, which is 40 percent for the proposed alternative, and greater than one full SG 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 C2.21 04-073 4/30/2017 3rd / 3rd / 1RF21 5th Int / 3rd Period 2035 18 C2.22 04-074 4/28/2017 3rd / 3rd / 1RF21 5th Int / 3rd Period 2035 18 C1.30 04-075 10/6/2020 4th / 1st / 1RF23 6th Int / 1st Period 2039 19 C1.20 04-068 4/29/2017 3rd / 3rd / 1RF21 5th Int / 3rd Period 2035 18 B2.40 04-076 5/4/2017 3rd / 3rd / 1RF21 5th Int / 3rd Period 2035 18 The NRC staff determined, through binomial statistics and Monte Carlo methods that a 25 percent sample of the total ASME Code required number of SGs would be an adequate performance monitoring sample over the subject alternative period. In the context of this SG submittal this would lead to a sample of 0.25x3 = 0.75 SGs. However, at least one SG 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 SG worth of examinations. Consequently, the 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 reasonably continuous source of data 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 appropriate amount of data.
As part of the proposed performance monitoring plan in the supplement dated September 24, 2024, the licensee described actions they 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 industrywide 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 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-24-02 for Entergy through the end of the current licensed operating life for 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 Contributors: C. Parker, NRR D. Dijamco, NRR Date: February 20, 2025
- concurrence via email 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