ML23256A088
| ML23256A088 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire  |
| Issue date: | 09/25/2023 |
| From: | David Wrona Plant Licensing Branch II |
| To: | Gibby S Duke Energy |
| Jordan, N | |
| References | |
| EPID L-2023-LLR-0003 | |
| Download: ML23256A088 (1) | |
Text
September 25, 2023 Mr. Shawn Gibby Vice President Nuclear Engineering Duke Energy 526 South Church Street, EC-07H Charlotte, NC 28202
SUBJECT:
CATAWBA NUCLEAR STATION, UNIT NOS. 1 AND 2; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1; MCGUIRE NUCLEAR STATION, UNIT NOS. 1 AND 2; OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3; AND H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF ALTERNATIVE TO STEAM GENERATOR WELDS (EPID L-2023-LLR-0003)
Dear Mr. Gibby:
By letter dated January 23, 2023, as supplemented by letter dated July 20, 2023, Duke Energy Carolinas (Duke Energy, the licensee), submitted a request for the Catawba Nuclear Station, Units 1 and 2 (CNS); McGuire Nuclear Station, Units 1 and 2 (MNS); Oconee Nuclear Station, Units 1, 2, and 3 (ONS); H. B. Robinson Steam Electric Plant, Unit 2 (RNP); and Shearon Harris Nuclear Power Plant, Unit 1 (HNP) to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV 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 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 has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that Duke Energy has adequately addressed the requirements in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative for Duke Energy through the 5th inservice inspection (ISI) interval for CNS and HNP and through the 6th ISI interval for RNP, MNS, and ONS.
All other ASME BPV Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact Shawn Williams at (301) 415-1009 or by e-mail at Shawn.Williams@nrc.gov.
Sincerely, David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: 50-413, 50-414, 50-400, 50-369, 50-370, 50-269, 50-270, 50-287, and 50-261
Enclosure:
Safety Evaluation cc: Listserv David J.
Wrona Digitally signed by David J. Wrona Date: 2023.09.25 16:15:16 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSERVICE INSPECTION INTERVAL EXTENSION FOR STEAM GENERATOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT 2 DOCKET NOS. 50-413, 50-414, 50-400, 50-369, 50-370, 50-269, 50-270, 50-287, AND 50-261
1.0 INTRODUCTION
By letter dated January 23, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23023A093), as supplemented by letter dated July 20, 2023 (ML23201A140), Duke Energy Carolinas (Duke Energy, the licensee), submitted a request for the Catawba Nuclear Station, Units 1 and 2 (CNS1 and CNS2); McGuire Nuclear Station, Units 1 and 2 (MNS1 and MNS2); Oconee Nuclear Station, Units 1, 2, and 3 (ONS1, ONS2, and ONS3); H. B. Robinson Steam electric Plant, Unit 2 (RNP); and Shearon Harris Nuclear Power Plant, Unit 1 (HNP) to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV 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 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 proposed alternative request for CNS1 and 2; MNS1 and 2; ONS1, 2, and 3; RNP; and HNP as a plant-specific alternative.
2.0 REGULATORY EVALUATION
The SG pressure-retaining welds and SG full penetration welded nozzles at the subject plants 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).
The regulations in 10 CFR 50.55a(g)(4) state, in part, 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) state, in part, that alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a may be used when authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Proposed Alternative
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Applicable Code Edition and Addenda===
The applicable ISI interval and associated codes of record for the subject plants are summarized in Table 1.
Table 1: Section XI Codes of Record for Subject Plants Plant ISI Interval ASME Section XI Code Edition/Addenda Current Interval Start Date Current Interval End Date CNS1 and CNS2 Fourth 2007 Edition through 2008 Addenda 08/19/2015 12/06/2024 (Unit 1) 02/24/2026 (Unit 2)
MNS1 Fifth 2007 Edition through 2008 Addenda 12/01/2021 11/30/2031 MNS2 Fourth 2007 Edition through 2008 Addenda 07/15/2014 02/29/2024 ONS1, ONS2, and ONS3 Fifth 2007 Edition through 2008 Addenda 07/15/2014 07/15/2024 RNP Fifth 2007 Edition through 2008 Addenda 07/21/2012 02/19/2023 HNP Fourth 2007 Edition through 2008 Addenda 09/09/2017 09/08/2027 American Society of Mechanical Engineers (ASME) Code Components Affected ASME Code Class:
Class 1 and 2 Examination Category:
B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels C-A, Pressure Retaining Welds in Pressure Vessels C-B, Pressure Retaining Nozzle Welds in Pressure Vessels Item Numbers:
B2.40 for SG vessel primary side, tubesheet-to-head welds C1.10, C1.20, and C1.30 for SG vessel secondary side welds C2.21 for the SG feedwater (FW) and main steam (MS) nozzle-to-vessel welds C2.22 for the SG FW and MS nozzle inside radius (NIR) sections Component IDs:
The nine tables in Section 1 of Enclosure 1 to the licensees submittal lists the component identifications (IDs) affected for each subject plant.
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 ASME Code,Section XI, Table IWB-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval. As noted in Table IWB-2500-1 for Examination Category B-B, 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 ASME Code,Section XI, Table IWC-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval. As noted in Table IWC-2500-1 for Examination Categories C-A and C-B, 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. C1.10, Shell Circumferential Welds 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-B, Item No. C2.21, Nozzle-to-Shell Welds Examination Category C-B, Item No. C2.22, Nozzle Inside Radius 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.
Reason for Proposed Alternative In Section 4.0 of Enclosure 1 to the submittal, 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 SE.
EPRI Technical Report 3002015906, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, 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 1.245, Preparing Probabilistic Fracture Mechanics Submittals (ML21334A158). Based on the conclusions of the two reports, the licensee requested an alternative to the ASME Section XI examination requirements for the subject welds.
The NRC staff noted that the EPRI reports were not submitted or reviewed as a topical report.
The NRC staff reviewed the proposed alternative request for the subject plants 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 Basis for Use In Section 5.0 of Enclosure 1 to the submittal, the licensee described the proposed alternative for each plant, as follows.
For CNS1, the licensee requested to defer the examinations for Item Numbers B.2.40, C1.20, C1.30, C2.21, and C2.22 through the end of the 5th inspection interval. The licensee stated that the CNS1 SGs were replaced in 1996. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 2nd period of the current 4th ISI interval. The licensee further explained that the proposed alternative leads to a maximum period of 19 years, 10 months, 10 days from the end of the 3rd ISI interval.
For CNS2, the licensee requested to defer the examinations for Item Numbers B2.40, C1.10, C1.20, C1.30, C2.21, and C2.22 through the end of the 5th ISI interval. The licensee stated that CNS2 is still using the originally-installed SGs. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 2nd period of the current 4th ISI interval. The licensee further explained that the proposed alternative leads to a period of 20 years from the end of the 3rd ISI interval.
For MNS1 and 2, the licensee requested to defer the examinations for Item Numbers B2.40, C1.20, C1.30, C2.21, and C2.22 through the end of the 5th ISI interval. The licensee stated that the MNS1 and 2 SGs were replaced in 1997. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 2nd period of the 4th ISI interval. The licensee further explained that the proposed alternative leads to a period of 20 years from the end of the 4th ISI interval.
For ONS1, 2, and 3, the licensee requested to defer the examinations for Item Numbers B2.40, C1.30, and C2.21 through the end of the 6th ISI interval. The licensee stated that ONS1, ONS2, and ONS3 SGs were replaced in 2003 and 2004. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 5th ISI interval.
For HNP, the licensee requested to defer the examinations for Item Numbers B.2.40, C1.20, C1.30, C2.21, and C2.22 through the end of the 5th inspection interval. The licensee stated that the HNP SGs were replaced in 2001. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 2nd period of the current 4th ISI interval. The licensee clarified that only the Item Number C2.21 examinations were performed during the 4th ISI interval. The licensee further explained that the proposed alternative leads to a period of 19 years, 7 months, 23 days from the end of the 3rd ISI interval.
For RNP, the licensee requested to defer the examinations for Item Numbers B.2.40, C1.10, C1.20, C1.30, C2.21, and C2.22 through the end of the 6th inspection interval. The licensee stated that the RNP SGs were replaced in 1984. According to the licensee, the subject SG welds and nozzles received the required PSI examinations followed by ISI examinations through the 5th ISI interval.
Figure 1 demonstrates the licensees proposed alternative from Figure 1-1 of the supplement.
Additionally, for each subject unit, Table 1-2 of the supplement shows the length of time from the last inspection to end of proposed alternative for that unit.
Figure 1: Licensees Proposed Alternative and Performance Monitoring Plan Duration of Proposed Alternative The licensee requested to apply the proposed alternative for the remainder of the current fourth 10-year ISI interval through the end of the fifth 10-year ISI interval for CNS1 and 2, and HNP.
The licensee requested to apply the proposed alternative for the fifth and sixth 10-year intervals for MNS1 and 2. The licensee requested to apply the proposed alternative for the sixth 10-year ISI interval for ONS1, 2, and 3, and RNP.
Basis for Proposed Alternative In Section 5.0 of Enclosure 1 to the 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 basis for proposed alternative for the SG ASME Code Examination Categories B-B and C-A welds. EPRI report 14590 was used as basis for proposed alternative for the SG ASME Code Examination Category C-B welds and NIR.
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 NRC 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 NRC 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 Duke Energy review, the NRC staff considered the referenced information and focused on the plant-specific application of the EPRI reports for the subject Duke Energy units. Using a risk-informed approach, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.
3.2
NRC Staff Evaluation
3.2.1 Degradation Mechanisms The NRC staff reviewed the submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to the subject Duke Energy units to merit additional consideration. The NRC staff found no evidence of conditions at the subject Duke Energy units 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 NRC staff reviewed the materials, stress states, and consistency of 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 NRC staff finds that consideration of additional degradation mechanisms beyond those from the EPRI reports is not necessary.
3.2.2 PFM Analysis The NRC staff confirmed that the PFM analysis referenced by the licensee for the Duke Energy submittal is consistent with the approach taken in the technical arguments presented in the Millstone and Vogtle submittals and explicitly referenced in the Duke Energy request. The original review of this approach is documented in the Millstone and Vogtle SEs. The NRC staff reviewed the application of this approach, as proposed in the Duke Energy request, and determined that the PFM analysis is consistent with the previously approved precedence in the Millstone and Vogtle submittals. Therefore, the NRC staff finds the proposed PFM analysis to be appropriate for this application for Duke Energy.
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 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 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), August 2007 (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 NRC 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 Duke Energy 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 sub-population which can be considered the principal system risk for integrity). The NRC 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 criteria such as 1x10-6 per year for individual welds may not be appropriate generically, but based on the discussion above, the NRC staff finds the application of this criterion acceptable for this plant-specific review for the SG welds and NIR for Duke Energy.
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 NRC staff accepted in proprietary report BWRVIP-05, BWR [Boiling Water Reactor] Vessel and Internals Project: 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. RG 1.154 was later withdrawn in 2011. The NRC staff also noted that the BWR Vessel and Internals Project topical reports included substantive inspection aspects that were critical to the NRCs findings.
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 the Duke Energy plant-specific alternative request.
3.2.3 Parameters Most Significant to PFM Results The NRC staff reviewed the submittal for plant-specific aspects of the Duke Energy alternative request that may diverge from the Millstone and Vogtle submittals, as explicitly referenced in the Duke Energy request, concerning parameters most significant to PFM results. The NRC staff confirmed that the review conclusions in the Millstone and Vogtle SEs applied to the Duke Energy submittal and found that the parameters most significant to PFM results would be the same and consistent with the NRC staffs reviews documented in the Millstone and Vogtle SEs, and consequently the approach taken in those reviews appropriately applies to the current review for Duke Energy.
As discussed in the Millstone and Vogtle SEs, the sensitivity analysis (SA), sensitivity study (SS), and the NRC staffs observations on the 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 NRC staff discussed and closely evaluated each in the next five sections of this SE. The NRC 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 Attachments 1 through 6 of the submittal, 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 Duke Energy. The licensee stated that most plant-specific criteria as specified in the EPRI reports 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 Duke Energy. 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 Duke Energy plant-specific alternative request. Specifically, the radius-to-thickness (R/t) ratios of the requested Duke Energy components, provided in Tables 1 and 2 of Enclosure 1 to the submittal, are either bounded by the R/t ratios analyzed in the EPRI reports or by the SS on stress in the EPRI reports.
Specifically, the RNP main steam nozzle R/t is within a factor of two of the recommended R/t (see Table 2 of Enclosure 1 of the licensees submittal), which was analyzed by the stress multiplier of 2.2 considered in Table 8-15 of EPRI report 14590. To verify the dominance of the R/t ratio, the NRC staff reviewed the through-wall stress distributions in Section 7 of the EPRI reports to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. For some of the SG shell welds modeled in EPRI report 15906, the NRC noted that the thermal stress is also potentially high, as discussed in the Millstone SE. However, because thickness is the controlling parameter for thermal stress (the lower the thickness, the lower the thermal stress), the NRC staff determined that EPRI report 15906 would still be adequate for the corresponding welds for the subject plants because the thickness values of the Duke Energy SG vessels, as shown in Table 1 of Enclosure 1 to the submittal, are less than those for the SG model analyzed in EPRI report 15906 except for CNS2 which would be adequately bounded by the SS on stress in the report. Accordingly, the NRC staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for the Duke Energy plant-specific alternative request.
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 B&PV Code,Section XI, Nonmandatory Appendix G, paragraph G-2110. The licensee addressed these criteria in Tables 1-1, 2-1, 3-1, 4-1, 5-1, and 6-1 of the Duke Energy submittal. The licensee stated that the materials of construction for the SG components are as reported in Table 2.
Table 2: Materials of Construction Unit SG Component Material CNS1 vessel heads SA-508, Class 3 vessel shell SA-533 Type B, Class 1 tubesheet SA-508 Class 3 feedwater nozzles SA-508 Class 3 main steam nozzles N/A CNS2 vessel heads SA-533 Grade A, Class 2 vessel shell SA-533 Grade A, Class 2 tubesheet SA-508 Class 2a feedwater nozzles SA-508 Class 2a main steam nozzles SA-508 Class 2a MNS1 and 2 vessel heads SA-508 Class 3 vessel shell SA-533 Type B, Class 1 tubesheet SA-508 Class 3 feedwater nozzles SA-508 Class 3 main steam nozzles N/A ONS1/2/3 vessel heads SA-508 Class 3a vessel shell SA-508 Class 3a tubesheet SA-508 Class 3a feedwater nozzles N/A main steam nozzles SA-508 Class 3a HNP vessel heads SA-508 Class 3a vessel shell SA-508 Class 3a tubesheet SA-508 Class 3a feedwater nozzles SA-508 Class 3a main steam nozzles N/A RNP vessel heads SA-302 Grade B vessel shell SA-533 Grade A Class 2 tubesheet SA-508 Class 2a feedwater nozzles SA-508 Class 2a main steam nozzles SA-533 Grade A Class 2 The NRC staff verified that these materials conform with ASME Code Section XI, Nonmandatory Appendix G. Therefore, the NRC staff finds that the materials for Duke Energy meet the material applicability criterion.
Tables 1-1, 2-1, 3-1, 4-1, 5-1, and 6-1 of the submittal state that the Duke Energy SG shell and nozzles meet the applicability criteria in EPRI reports 14590 and 15906 regarding weld and nozzle configuration, attached piping line size, and thermal sleeve. However, the licensee stated in Table 4-1 of the submittal that the upper and lower SG heads for ONS1, 2, and 3 undergo a diameter reduction, such that the diameter of the head is less than the diameter of the shell. The licensee noted that this geometry is consistent with the analyzed B&W design in EPRI report 15906 and consequently bounded by those analyses. The NRC staff confirmed that the ONS SGs were manufactured by B&W, and that their geometry is sufficiently consistent with the B&W case analyzed (see Figure 4-1 of the submittal in comparison with Figure 4-3 of EPRI report 15906). The NRC staff noted that based on this comparison, the diameter of the SG lower head for the B&W design in Table 9-2 of EPRI report 15906 is incorrectly reported and should be a lower value. The NRC staff further noted that the correct SG lower head diameter was analyzed in EPRI report 15906. Therefore, the NRC staff finds that the geometry of the ONS1, 2, and 3 SG heads are bounded by the EPRI analysis. Overall, the NRC staff reviewed the licensees information against the applicability criteria and finds that the Duke Energy SGs meet the applicability criteria described in the EPRI reports.
3.2.4.2 Selection of Transients In Section 5.2 of EPRI reports 14590 and 15906, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to the SG shell and SG nozzles. EPRI developed a list of transients for analysis applicable to the SG shell and SG nozzles analyzed in the report, based on transients that have the largest temperature and pressure variations.
The NRC staff evaluated the transient selection in the EPRI reports in detail, as discussed in the Millstone and Vogtle SEs. The NRC staff confirmed that the applicable aspects of the transients discussed in those SEs apply equally to this review for Duke Energy. The NRC 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 Duke Energy 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 NRC staff then compared the analysis in the EPRI reports to plant-specific information provided in the licensees submittal.
In Tables 1-2 through 1-4 (CNS1), Tables 2-2 through 2-4 (CNS2), Tables 3-2 through 3-4 (MNS1/2), Tables 4-2 through 4-4 (ONS1/2/3), Tables 5-2 through 5-4 (HNP), and Tables 6-2 through 6-4 (RNP) of the 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 Duke Energy. The NRC staff reviewed these tables and confirmed that the Duke Energy SG shells and SG nozzles are bounded by the criteria in the EPRI reports. The NRC staff noted that there were minor differences in temperatures and pressures. However, the NRC 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 1-4, 2-4, 3-4, 4-4, 5-4, and 6-4 of the submittal are substantially below the number of cycles assumed in the analysis. In the supplement, the licensee confirmed that the 60 cycles of Loss of Power transient analyzed in EPRI report 14590 is bounding for RNP. Therefore, the NRC staff finds that the transients assumed in the EPRI analysis appropriately bound the subject plants.
In the analyses in the EPRI reports there were no separate test conditions included in the transient selection. The licensee stated on page 15 of 29 of Enclosure 1 to the submittal that pressure tests (i.e., system leakage tests) for Duke Energy 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 Duke Energy meets the transient applicability criteria in the EPRI reports. Therefore, the analyzed transient loads for the requested SG components at Duke Energy 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 NRC staff confirmed that no Duke Energy plant-specific aspects of this submittal warranted additional consideration, noting in particular (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 NRC staff finds that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on the probability of leakage or rupture beyond the studies documented in 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 Duke Energy.
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 NRC staff confirmed that no Duke Energy plant-specific aspects of this application warranted further review. Based on this, the NRC staff determined that the pressure and thermal stresses calculated through FEA in the EPRI reports are acceptable for referencing for the requested SG welds and NIR of Duke Energy.
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 Duke Energy submittal without further plant-specific considerations. As discussed in Section 3.2.4 of this SE, Duke Energy meets the material criteria in the EPRI reports, and thus the NRC staff determined that the assumed fracture toughness parameters above are applicable to Duke Energy.
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 Duke Energy, the NRC staff finds the fracture toughness model in the referenced EPRI reports acceptable for the requested SG welds and NIR of Duke Energy.
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 NRC staff confirmed that the evaluation documented in the Vogtle SE applies to the Duke Energy 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, Duke Energy 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 Duke Energy. The flaw density in the SG welds 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 SG welds, the NRC 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 Duke Energy, the NRC staff finds the appropriate flaw density has been considered, and therefore acceptable, for the requested SG welds and NIR of Duke Energy.
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 Millstone and Vogtle SEs. The NRC staff confirmed that no plant-specific aspects of the Duke Energy submittal warranted further review with regards to FCG rate. Based on the discussions referenced above, the NRC 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 NIR of Duke Energy.
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 Duke Energy in the following tables in the submittal: Tables 1-5 (CNS1), 2-5 (CNS2), 3-5 and 3-6 (MNS1/2), 4-5 through 4-7 (ONS1/2/3), 5-5 (HNP), and 6-5 (RNP). These tables indicate that there were no unacceptable indications found during these examinations. The licensee stated in Section 5.0 of Enclosure 1 to the submittal that some SGs of the subject Duke units have been replaced. Given the implementation of ISI in the PFM analyses in the EPRI reports, as the NRC staff explained in the Millstone and Vogtle SEs, the NRC 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. The ISI examination scenarios considered are shown in Table 3 of this SE.
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) 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 also noted an examination coverage of about 70% at HNP for a B2.40 weld, one of the limiting locations analyzed in EPRI report 15906. This situation is similar to that discussed in the Millstone SE (Section 11.1), in which the NRC staff determined that the limiting location that had an examination coverage of 50% would have a probability of rupture less than the acceptance criterion of 1x10-6 per year. Therefore, the NRC staff determined that the limiting location at HNP with examination coverage of 70% would have a probability of rupture less than the acceptance criterion of 1x10-6 per year.
Based on this discussion, the NRC staff finds the Duke Energy 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 Duke Energy, the NRC staff finds that the PFM analyses of EPRI reports 14590 and 15906 adequately represent the requested components for Duke Energy 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 NRC 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 NRC staff confirmed that no plant-specific aspects of the submittal warranted further review. Based on the discussion referenced above, the NRC staff finds that the submittal is acceptable with regards to these modeling aspects used in the EPRI reports, and therefore, is acceptable for the requested SG welds and NIR of Duke Energy.
3.2.10 PFM Results Relevant to Proposed Alternative In Section 5.0 of Enclosure 1 to the 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 NRC 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.
Duke Energy evaluated the PFM scenarios on pages 16-18 of Enclosure 1 of the submittal in order to determine which scenarios best represent the inspection history of the subject nuclear plants. This evaluation depends upon whether the plants SGs were replaced, the date of the PSI examination, and the dates of subsequent ISI examinations. The scenarios identified by the licensee are found in Table 3.
Table 3: Examination Scenarios in the PFM Study Unit PFM Scenario CNS1 PSI+10+20+50 CNS2 PSI+10+20+30+60 MNS1 PSI+10+20+50 MNS2 PSI+10+20+50 ONS1/2/3 PSI+10+20+50 HNP PSI+10+20+50 RNP PSI+10+20+30+40+70 Duke Energy stated that the limiting scenario for this alternative request is PSI+10+20+50. The licensee then performed five sensitivity study evaluations with this limiting scenario since this scenario was not specifically considered in the EPRI reports, using the following cases from the reports: Case ID FEW-P1N for the inside radius of the feedwater nozzle, Case ID FEW-P3A for the feedwater nozzle-to-shell weld, Case ID SGB-P1N for the main steam nozzle inside radius, Case ID SGB-P3A for the main steam nozzle-to-shell weld, and Case ID SGPTH-P4A for the SG shell welds. The licensee reported that these analysis cases resulted in PoFs less than the chosen acceptance criterion of 1x10-6 per year.
3.2.11 Performance Monitoring Performance monitoring, such as inservice inspection 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 Duke Energy would have resulted in a significant amount of time before another examination was performed on several 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 remain valid throughout the period of the proposed alternative (see RAI-1 in the July 20, 2023, supplement).
In response to RAI-1, the licensee described 7 examinations to be performed throughout the period of the proposed alternative. The proposed examinations will be performed at 5 of the 9 units covered by the proposed alternative (i.e., CNS1 and 2, MNS1 and 2, and HNP). Each of the ASME item numbers covered by the proposed alternative (i.e., B.2.40, C1.10, C1.20, C.1.30, C.2.21, and C2.22) will be examined as part of the proposed performance monitoring plan. The licensee also stated that for the 6th ISI intervals of CNS1 and 2 and HNP, the ASME Code,Section XI required examination will resume. The licensees proposed performance monitoring plan schedule is shown in Figure 1-1 of Enclosure 1 to the July 20, 2023, supplement, reproduced in this SE as Figure 1. The ASME Code Section XI requirement is to perform 29 examinations in this time period. The licensee stated that performing 7 of the 29 required examinations is 7/29 = 24.1% for one interval.
The NRC staff reviewed the proposed performance monitoring plan in terms of SGs, rather than in terms of number of required Item No. examinations. ASME Code Section XI requires that one SG be examined per unit per ISI interval. Accordingly, Table 4 shows the calculation for number of total ASME Code required SGs for the licensees proposed alternative.
Table 4: Calculation of Total ASME Code Required SGs Site
- of units
units x intervals Catawba 2
2 4
McGuire 2
2 4
Harris 1
2 2
Oconee 3
1 3
Robinson 1
1 1
Total 14 The NRC staff determined, through binomial statistics and Monte Carlo methods, that a 25%
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 and similar SG submittals referencing the EPRI reports, this leads to a sample of 0.25 x 14 = 3.5 SGs. The SG equivalents proposed in the licensees performance monitoring plan is shown in Table 5.
Table 5: Number of SG Equivalent Exams Unit
- of Section XI exams
- of performance monitoring exams SG Equivalents =
PM exams / required exams Catawba 1 5
2 0.4 Catawba 2 9
2 0.222 McGuire 1 5
1 0.2 McGuire 2 5
1 0.2 Harris 5
1 0.2 Total 1.22 In addition to the SG equivalents calculated in Table 5, CNS1 and 2 and HNP will resume ASME Code,Section XI required exams for the 6th ISI interval for those respective plants. This leads to a total SG equivalents of 1.22 + 3 = 4 for the subject time period, which meets the NRC staff position of at least 25%. 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 reasonably continuous source of data supporting the characteristics of acceptable performance monitoring. Specifically, data would continue to become available on a cadence reasonably commensurate with ASME Code requirements, but on a fleet basis rather than an individual unit basis. Based on the proposed examinations during the alternative periods, and periods in which unmodified ASME Code inspections requirements would be in force, the NRC staff finds that the examinations proposed in the performance monitoring plan will provide an appropriately continuous stream of data.
As part of the proposed performance monitoring plan in the supplement dated July 20, 2023, the licensee described actions they would take in the event that 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. Furthermore, the licensee stated that detected indications would result in additional examinations at the CNS1 and 2, MNS1 and 2, and HNP plants within the first or second refueling outage of discovering the indication. The licensee stated that domestic and international operating experience would be entered into the Duke Energy Corrective Action Program to determine if additional examinations are required.
In addition, in the submittal the licensee discussed system leakage tests as providing assurance of safety for the proposed alternative. The NRC staff noted that the visual examinations performed during system leakage tests may not directly detect the presence or extent of degradation; may not provide direct detection of aging effects prior to potential loss of structure or intended function; and do not provide sufficient validating data necessary to confirm the modeling of degradation behavior in the subject SG welds. However, the NRC staff noted that leakage tests provide complementary additional performance monitoring to the ISI examinations. This additional assurance increases confidence that the proposed quantity of examinations, in concert with other on-going activities, will provide an acceptable level of performance monitoring for the subject SG components.
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 determined that the licensees proposed alternative as discussed above 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 for Duke Energy through the 5th ISI interval for CNS1 and 2 and HNP and through the 6th ISI interval for RNP; MNS1 and 2; and ONS1, 2, and 3.
All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: M. Benson, NRR D. Dijamco, NRR D. Widrevitz, NRR Date: September 25, 2023
ML23256A088 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-2/LA NRR/DNRL/NVIB/BC NAME SWilliams RButler ABuford DATE 09/22/2023 09/19/2023 09/08/2023 OFFICE DORL/LPL2-2/BC NAME DWrona DATE 09/25/2023