ML25014A286
| ML25014A286 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/21/2025 |
| From: | Gonzales H NRC/NRR/DORL/LPL1 |
| To: | Carr E Constellation Energy Generation |
| References | |
| EPID L-2024-LLR-0023 | |
| Download: ML25014A286 (1) | |
Text
January 21, 2025 Eric S. Carr Chief Nuclear Officer Dominion Energy Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 3 - ALTERNATIVE REQUEST TO DEFER ASME CODE SECTION XI INSERVICE INSPECTION EXAMINATIONS FOR PRESSURIZER AND STEAM GENERATOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES (EPID L-2024-LLR-0023)
Dear Eric Carr:
By letter dated March 22, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24086A480), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted Alternative Request IR-4-14 to the U.S. Nuclear Regulatory Commission (NRC) for authorization of an alternative to the inservice inspection (ISI) requirements American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Table IWB-2500-1, Examination Category 8-8 and 8-D and Table IWC-2500-1, Examination Category C-A and C-8, component examinations for Millstone Power Station Unit 3 (Millstone, Unit 3), in association with the remaining two 10-year ISI intervals from the last examination performed for each item.
Specifically, pursuant to subparagraph (1) in paragraph (z), Alternatives to codes and standards requirements, of Section 55a, Codes and standards, in Part 50, Domestic Licensing of Production and Utilization Facilities, to Title 10, of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to defer selected pressurizer and steam generator weld examinations through the fifth and sixth 10-year ISI intervals.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, 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 Millstone, Unit 3, to increase the ASME Code,Section XI, 10-year ISI interval for the subject welds (and associated nozzle inside radius where applicable), thereby deferring the examinations for those welds and the nozzle inside radius sections for the fifth and sixth 10-year ISI intervals. The subject welds will be reexamined prior to the end of the current 60-year operating license for Millstone, Unit 3, which expires on November 25, 2045. The sixth 10-year ISI interval is projected to end in February of 2049.
All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested, and granted or authorized (as appropriate), in the subject request remain applicable.
E. Carr If you have any questions, please contact the Project Manager, Richard Guzman, at 301-415-1030 or Richard.Guzman@nrc.gov.
Sincerely, Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosure:
Safety Evaluation cc: Listserv HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.01.21 16:52:23 -05'00'
Enclosure STAFF EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST IR-4-14 TO DEFER ASME CODE SECTION XI EXAMINATION OF PRESSURIZER AND STEAM GENERATOR WELDS AND NOZZLES DOMINION ENERGY NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 DOCKET NO. 50-423
1.0 INTRODUCTION
By letter dated March 22, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24086A480), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted a request for the Millstone Power Station (Millstone) Unit 3, to the U.S.
Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to defer certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI examinations.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed to defer the ASME Code,Section XI, 10-year inservice inspection (ISI) interval for the requested components, thereby deferring the examinations for the respective components for two 10-year ISI intervals from the last examination performed for each item, on the basis that the proposed alternative provides an acceptable level of quality and safety. In its submittal, the licensee requested an alternative on the basis of plant-specific applicability of three technical reports prepared by the Electric Power Research Institute (EPRI). The NRC staff reviewed the proposed alternative request for Millstone, Unit 3, as a plant-specific alternative.
The NRC did not review the EPRI reports for generic use, and this approval does not extend beyond the Millstone, Unit 3, plant-specific authorization.
2.0 REGULATORY EVALUATION
The pressurizer (PZR) and steam generator (SG) pressure-retaining welds and SG full penetration welded nozzles at Millstone, Unit 3, are ASME Code Class 1 and 2 components, whose ISIs are performed in accordance with Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME Code and applicable edition and addenda, 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.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the alternative and the NRC staff to authorize it.
3.0 LICENSEES PROPOSED ALTERNATIVE 3.1
Applicable Code Edition and Addenda
Millstone, Unit 3, is currently in the fourth 10-year ISI interval (February 23, 2019, to February 22, 2029), and the ASME Code of record for this ISI interval is the 2013 Edition of the ASME Code,Section XI.
3.2 ASME Code Components Affected
ASME Code Class:
Section XI, Class 1 and 2 Examination Category:
B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels B-D, Full Penetration Welded Nozzles Vessels C-A, Pressure Retaining Welds in Pressure Vessels C-B, Pressure Retaining Nozzle Welds in Pressure Vessels Item Numbers:
B2.11 - PZR, shell-to-head welds, circumferential B2.12 - PZR, shell-to-head welds, longitudinal B2.40 - SG (primary side), tubesheet-to-head welds B3.110 - PZR, nozzle-to-vessel welds B3.130 - SG, nozzle-to-vessel welds C1.10 - Shell circumferential welds C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell welds C2.21 - Nozzle-to-shell (nozzle-to-head/nozzle-to-nozzle) welds C2.22 - Nozzle inside radius (NIR) sections Component IDs:
Table 2 of Attachment 1 to the submittal lists the component identifications (IDs) affected.
3.3 ASME Code Requirements for Which Alternative Is Requested For ASME Code Class 1 welds in the PZR and SG listed below, 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:
Examination Category B-B, Item No. B2.11, PZR, shell-to-head welds, circumferential Examination Category B-B, Item No. B2.12, PZR, shell-to-head welds, longitudinal Examination Category B-D, Item No. B3.110, PZR, nozzle-to-vessel welds Examination Category B-D, Item No. B3.130, SG, nozzle-to-vessel welds For ASME Code Class 1 welds in the SG listed below, 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 Examination Category C-B, Item No. C2.22, NIR sections 3.4 Reason for Proposed Alternative In Section 4.0 of Attachment 1 to its submittal, the licensee stated that the EPRI performed assessments in the following non-proprietary reports of the basis for the ASME Code,Section XI examination requirements for the PZR and SG welds and components for the requested alternative.
EPRI Technical Report 3002015905, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, 2019 (hereafter referred to as EPRI Report 3002015905, ML21021A271).
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 (hereafter referred to as EPRI Report 3002015906, 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 (hereafter referred to as EPRI Report 3002014590, 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 that the reports concluded that the current ASME Code,Section XI ISI interval of 10 years can be increased significantly with no impact to plant safety. Based on the conclusions of the three EPRI reports, the licensee is requesting an alternative to the 10-year ISI interval for the subject welds.
The NRC staff noted that the EPRI reports were not submitted or reviewed as topical reports.
The NRC staff reviewed the proposed alternative request for Millstone, Unit 3, as a plant-specific alternative. The NRC did not review the EPRI reports for generic use, and this review does not extend beyond the Millstone, Unit 3, plant-specific authorization.
3.5 Proposed Alternative and Basis For Use In Section 5.0 of Attachment 1 to its submittal, the licensee stated that the proposed alternative increases the ISI interval for these examination items from the current ASME Code,Section XI, 10-year requirement thereby deferring examinations for two 10-year ISI intervals from the last examination performed for each item.
3.6 Duration of Proposed Alternative The licensee requested to apply the proposed alternative, thereby deferring examinations for two 10-year ISI intervals from the last examination performed for each item. The subject welds will be reexamined prior to the end of the current 60-year operating license for Millstone, Unit 3, which expires on November 25, 2045. The sixth 10-year ISI interval is projected to end in February of 2049.
3.7 Basis for Proposed Alternative In Section 5.0 of Attachment 1 to its submittal, the licensee discussed the key aspects of the technical basis in the EPRI report and its applicability to Millstone, Unit 3. EPRI Report 3002015905 was used as basis for proposed alternative for the PZR ASME Code Examination Categories B-B and B-D welds. EPRI Report 3002015906 was used as basis for proposed alternative for the SG ASME Code Examination Categories B-B, B-D, and C-A welds. EPRI Report 3002014590 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 3002015905 and EPRI Report 3002015906, and Section 8.2 of EPRI Report 3002014590, and verifying whether the DFM and PFM analyses in the reports support the proposed alternative. The licensee cited NRC-approved precedents for its request that were based on EPRI Reports 3002015905, 3002015906, and 3002014590. These precedents included the submittals for Salem Generating Station (Salem), Units 1 and 2 (ML20218A587), Millstone Power Station (Millstone), Unit 2 (ML20198M682), and Vogtle Electric Generating Plant, Units 1 and 2 (ML19347B105). The licensee referenced applicable portions of the technical arguments from these submittals. The NRC staff documented its review of these applications in the associated plant-specific safety evaluations (SEs) (Salem, Units 1 and 2 ML21145A189), Millstone, Unit 2 (ML21167A355), and Vogtle (ML20352A155), hereafter referred as the Salem, Millstone Unit 2, and Vogtle SEs, respectively). For the Millstone, Unit 3, review, the NRC staff considered the information referenced and focused on the plant-specific application of the EPRI reports for Millstone, Unit 3.
Consistent with the key principles of the NRC risk-informed approach, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.
4.0 NRC STAFF EVALUATION 4.1 Degradation Mechanism In Section 5.0 of Attachment 1 to its submittal, the licensee stated, in part, that:
The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the PZR and SG welds and components covered in this request. Therefore, only those fatigue related mechanisms considered in the PFM and DFM evaluations in References
[1-2], [1-3], and [1-4] are applicable to the components in this request.
The NRC staff reviewed the submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to Millstone, Unit 3, to merit additional consideration. Such circumstances pertain to materials of the subject components, stress states, and reactor coolant environment. The NRC staff found no evidence of circumstances at Millstone, Unit 3, that would require consideration of a unique degradation mechanism beyond application of EPRI Reports 3002015905, 3002015906, and 3002014590.
Specifically, the NRC staff reviewed the materials, stress states, and chemical environment (i.e.,
reactor coolant) of the subject PZR and 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.
4.2 PFM Analysis In Section 5.0 of Attachment 1 to its submittal, the licensee stated, in part, that:
Finite element analyses (FEA) were performed in References [1-2], [1-3], and [1-4] to determine the stresses in the PZR and SG welds and components covered in this request. The finite element models used in References [1-2], [1-3], and [1-4] are consistent with the configurations for MPS3, therefore no new FEA model is required for the stress analysis of MPS3. The analyses were performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to MPS3 is demonstrated in Attachments 2 and 3 and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions contained in References [1-2], [1-3], and [1-4] are applicable to MPS3. In particular, the key geometric parameters used in the stress analyses in References [1-2], [1-3], and [1-4] are compared to those of MPS3, in Tables 5 and 6 for the PZR and Tables 7 and 8 for the SGs.
The licensee also stated, in part, that:
Flaw tolerance evaluations were performed in References [1-2], [1-3], and [1-4]
consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a PSI followed by subsequent ISI, the NRCs safety goal of 1.0x10-6 failures per year is met.
The NRC staff confirmed that the analysis provided by the licensee for the Millstone, Unit 3, submittal is consistent with the approach taken in the Salem, Millstone, Unit 2, and Vogtle precedents and explicitly referenced for plant-specific applicability in the Millstone, Unit 3, request. The NRC staff determined that the PFM analysis is consistent and, therefore, finds the proposed PFM analysis to be appropriate for this application for Millstone, Unit 3. The NRC staff evaluated the licensees submission about the impact of PSI on the PFM analysis in Section 4.10 of this SE.
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 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, would meet the guidance in Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256). 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 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 that is based on the reactor vessel TWCF criterion is reasonable for the requested PZR and SG welds and NIR of Millstone, Unit 3, because (a) the impact of a PZR or SG vessel failure would be less than the impact of a reactor vessel failure on overall risk (i.e., meaning contribution to core damage frequency due to a PZR or SG vessel failure would be less than the contribution due to a reactor vessel failure); (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, which means that failure of an individual weld is likely to lead to only a limited contribution to risk). The NRC staff further noted that comparing the probability of leakage to the same criterion of 1x10-6 failures per year is conservative because leakage is not failure. 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 PZR and SG welds and NIR for Millstone, Unit 3.
Based on the above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for the Millstone, Unit 3, plant-specific alternative request.
4.3 Parameters Most Significant to PFM Results In the following sections, the NRC staff reviewed the following parameters or aspects most significant to the PFM analysis: stress analysis, fracture toughness, flaw density, flaw crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage.
4.4 Stress Analysis 4.4.1 Selection of Components and Materials In Attachment 2 to its submittal, the licensee evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI Reports 3002015905, 3002015906, and 3002014590 to the subject PZR and SG welds and NIR of Millstone, Unit 3. These reports evaluated representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The reports also specified plant-specific applicability criteria with regards to component geometries, materials, and loading conditions, that must be evaluated and met by each plant to determine the applicability of the reports. The licensee stated that the plant-specific applicability criteria regarding component geometries and materials were met. The acceptability of meeting these criteria, however, depends on the acceptability of the component and material selection described in the EPRI reports, which the NRC staff evaluated below. The NRC staff evaluated the loading conditions (i.e., transient selection) criteria in Section 4.4.2 of this SE.
In Section 4 of EPRI Reports 3002015905, 3002015906, and 3002014590, EPRI discussed the variation among PZR, SG shell, and SG nozzle designs. EPRI used this information for finite element analyses (FEA, see Section 4.4.4 of this SE) to determine stresses in the analyzed components, which the licensee referenced for the corresponding PZR and SG components requested for Millstone, Unit 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 3002015905, 3002015906, and 3002014590, and finds that the PZR and SG configurations selected in the report for stress analysis are acceptable representatives for the corresponding PZR and SG components requested for the Millstone, Unit 3, plant-specific alternative request. Specifically, the radius-to-thickness (R/t) ratios of the requested Millstone, Unit 3, components, provided in Tables 5 through 8 of Attachment 1 to the submittal, are either bounded by the R/t ratios analyzed in the EPRI reports or such that they are bounded by the SS on stress in the EPRI reports. 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 3002015906, the NRC noted that the thermal stress is also potentially high.
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 3002015906 would still be adequate in providing reasonable assurance for the corresponding welds of Millstone, Unit 3, because the thickness values of the Millstone, Unit 3, SG vessel, as shown in Table 7 of Attachment 1 to the submittal, are less than those for the SG model analyzed in EPRI Report 3002015906. 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 Millstone, Unit 3, plant-specific alternative request.
Section 9.4 of EPRI Reports 3002015905, 3002015906, and 3002014590, 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 Table A1, Attachment 2 to the submittal for Millstone, Unit 3.
The materials of the Millstone, Unit 3, PZR and SGs relevant to the requested welds are SA-533, Grade A, Class 2; SA-533 Grade B, Class 1; SA-508, Class 2; SA-508, Class 2a, and SA-508, Class 3a. The NRC staff verified that these materials conform with ASME Code,Section XI, Paragraph G-2110. Therefore, the NRC staff finds that the materials for Millstone, Unit 3, meet the material applicability criterion.
Table A1, Attachment 2 to the submittal also states that the Millstone, Unit 3, PZR, and SG shell and nozzles meet the applicability criteria in EPRI Reports 3002015905, 3002015906, and 3002014590 regarding weld and nozzle configuration, attached piping line size, and thermal sleeve attachment. The NRC staff reviewed the licensees information against the applicability criteria and finds that the Millstone, Unit 3, PZR and SG meet the applicability criteria described in the EPRI reports.
Based on the above, the NRC staff finds that the licensee has made a plant-specific case that Millstone, Unit 3, meets the component geometry and materials applicability criteria in the EPRI reports. The analyzed geometries and materials are acceptable for the requested PZR and SG components at Millstone, Unit 3. The NRC concludes that the analyses acceptably model the subject Millstone, Unit 3, geometries and materials.
4.4.2 Selection of Transients In Attachment 2 to its submittal, the licensee evaluated the plant-specific applicability of the transients selected and analyzed in EPRI Reports 3002015905, 3002015906, and 3002014590 to the subject PZR and SG welds and NIR of Millstone, Unit 3. The licensee stated that the plant-specific applicability criteria regarding transients were met. The acceptability of meeting the criteria, however, depends on the acceptability of the transient selection described in the EPRI reports, which the NRC staff evaluated below.
In Section 5.2 of EPRI Reports 3002015905, 3002015906, and 3002014590, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to PZRs, SG shell, and SG nozzles. EPRI developed a list of transients for analysis applicable to the PZRs, 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 Salem, Millstone, Unit 2, and Vogtle SEs. The NRC staff confirmed that the applicable aspects of the transients discussed in those SEs apply equally to this review for Millstone, Unit 2. The NRC staff reviewed the discussion of transients in Section 5.2 of EPRI Reports 3002015905, 3002015906, and 3002014590, and determined that the transient selection defined in the reports are reasonable for the Millstone, Unit 3, plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG that are expected to occur in PWRs. The NRC staff then compared the analysis in the EPRI reports to plant-specific information provided in the licensees submittal.
In Tables A2, A3, A5, A6, A7 of Attachment 2 and Table A2 of Attachment 3 to the submittal, the licensee evaluated the plant-specific applicability of the transients selected in the EPRI Reports 3002015905, 3002015906, and 3002014590 to the PZR, SG shell, and SG nozzles of Millstone, Unit 3. The NRC staff reviewed the transient tables in Attachments 2 and 3 to the submittal, and confirmed that Millstone, Unit 3, is bounded by the criteria in the EPRI reports.
In the analyses in the EPRI reports there was no separate test conditions included in the transient selection. The licensee stated in Section 5.0 of Attachment 1 to the submittal that pressure tests for Millstone, Unit 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 above, the NRC staff finds that Millstone, Unit 3, meets the transient applicability criteria in the EPRI reports. The analyzed transient loads are acceptable for the requested PZR and SG components at Millstone, Unit 3. The NRC finds that the analyses acceptably model transients.
4.4.3 Other Operating Loads The NRC staff reviewed the application with regards to weld residual stress and clad residual stress. Weld residual stress and cladding stresses are addressed in EPRI Reports 3002015905, 3002015906, and 3002014590. The NRC staff documented the review of these aspects of the EPRI reports in the Salem, Millstone Unit 2, and Vogtle SEs. The NRC staff determined that no Millstone, Unit 3, plant-specific aspects of this submittal warranted additional consideration because of (1) the relatively low sensitivity of the EPRI results on residual stress (Table 8-14 of EPRI Report 3002015905, Table 8-12 of EPRI Report 3002015906, and Table 8-12 of EPRI Report 3002014590) 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 above, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested PZR and SG welds and NIR of Millstone, Unit 3. The NRC concludes the analyses acceptably bound other operating loads.
4.4.4 Finite Element Analysis The NRC staff reviewed the application with regards to FEA. FEA were conducted in EPRI Reports 3002015905, 3002015906, and 3002014590 as part of the stress analysis portion of the PFM analyses. The NRC staff documented its review in the Salem, Millstone, Unit 2, and Vogtle SEs. The NRC staff determined that no Millstone, Unit 3, plant-specific aspects of this application warranted further review because the FEA were performed with the representative component geometries, materials, and loading conditions discussed in Sections 4.4.1 and 4.4.2 of this SE for which the licensee provided plant-specific information and met the plant-specific criteria.
Based on the above, the NRC staff finds that the plant-specific Millstone, Unit 3, submittal is acceptable with regards to FEA.
4.5 Fracture Toughness In Attachment 2 to its submittal, the licensee stated that the materials of the subject Millstone, Unit 3, components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.
As discussed in Section 4.4.1 of this SE, the NRC staff verified that these materials conformed to the requirements of ASME Code,Section XI, Paragraph G-2110. In EPRI Reports 3002015905, 3002015906, and 3002014590, 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. The A-4200 fracture toughness curve refers to the same fracture toughness curve in ASME Code,Section XI, Paragraph G-2110. The NRC staff documented the review of the A-4200 fracture toughness value in the Salem, Millstone, Unit 2, and Vogtle SEs. In comparison, the NRC staff determined that the plant-specific Millstone, Unit 3, submittal is acceptable with regards to fracture toughness because the materials of the subject Millstone, Unit 3, components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.
4.6 Flaw Density In Attachment 1 to its submittal, the licensee stated that, per the Vogtle SE, a nozzle flaw density of 0.1 flaws per nozzle should have been used, and that the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10-6 failures per year. Further discussion of this topic as it relates to EPRI Report 3002014590 is contained in the Vogtle SE. The NRC staff noted that the flaw density of 0.1 flaws per nozzle is applicable to a specific plant so long as the component geometries and materials applicability criteria discussed in Section 4.4.1 of this SE are met. As discussed in that section of this SE, the licensee provided plant-specific information regarding the geometries and materials of the subject Millstone, Unit 3, components and met the applicability criteria. Based on the above, the NRC staff finds that the appropriate flaw density has been considered and is, therefore, acceptable for the requested SG NIR of Millstone, Unit 3.
4.7 FCG Rate The NRC staff reviewed the application with regards to FCG rate. The FCG rate used in EPRI Reports 3002015905, 3002015906, and 3002014590 is based on the ASME Code,Section XI, A-4300 FCG rate. The NRC staff documented its review in detail in the Salem, Millstone, Unit 2, and Salem SEs. The NRC staff noted that FCG rate depends on component material and reactor coolant environment. As discussed in Section 4.4.1 of this SE, the licensee provided plant-specific information regarding the materials of the subject Millstone, Unit 3, components and met the criteria for component materials. Per the ASME Code,Section XI, the A-4300 FCG rate may be used for light-water-cooled plants. Since Millstone, Unit 3, is a PWR, one of the two major types of light-water-cooled reactor designs, the NRC staff determined that the A-4300 FCG rate is appropriate for Millstone, Unit 3. Based on the above, the NRC staff finds that the plant-specific Millstone, Unit 3, submittal is acceptable with regards to FCG rate.
4.8 ISI Schedule and Examination Coverage In Attachment 4 to its submittal, the licensee provided information on the inspection history of the requested PZR and SG welds and NIR of Millstone, Unit 3, which consists of the ISI schedule and examination coverage. The licensee stated in Section 5.0 of Attachment 1 to the submittal that for the Millstone, Unit 3, PZR, preservice inspections (PSI) have been performed followed by ISI examinations over three complete 10-year ISI intervals. The licensee stated in Table A10 of Attachment 4 to its submittal that the Millstone, Unit 3, SG welds were subject to PSI followed by ISI examinations over three complete 10-year ISI intervals and some welds were inspected in the fourth 10-year ISI interval.
The licensee provided the inspection history of the requested PZR and SG welds and NIR of Millstone, Unit 3, in Attachment 4 to the submittal. The inspection history shows that examinations for the subject components were performed preservice and during the first, second, third, and to a limited extent in the fourth ISI intervals. The inspection history also shows that there is no evidence of unacceptable flaws in these components, which is consistent with other known operating experience and histories. The licensee noted that older examination reports did not report the examination coverage, and therefore, used the more recent examination coverage for the earlier examination coverage that was not documented, based on the consistency of the coverages throughout the examination history for the particular weld or component. The NRC staff determined that the licensees use of the more recent examination coverage for the earlier examination coverage that was not documented to be reasonable because (1) for the SG welds, the examination coverage values are relatively high (greater than 90 percent); and (2) for the PZR welds, probability of rupture is relatively insensitive to examination coverage. 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.
Specifically, the NRC staff noted one such instance in which there was relatively low examination coverage (55.5%) during earlier ISIs for four B3.130 welds (component IDs 03-003-SW-U and 03-003-SW-V in steam generator 1A and component IDs 03-004-SW-U and 03-004-SW-V in steam generator 1B). The NRC staff evaluated the potential impact of this examination coverage in earlier ISIs since Tables 8-32 and 8-33 of EPRI report 3002015906 show that the impact of a 50 percent coverage on item no. B3.130 increased the probability of rupture from 1.25x10-9 per year to 4.63x10-8 per year, which exceeds the acceptance criterion of 1x10-6 per year. Further, the 4.63x10-8 per year probability of rupture from the B3.130 Combustion Engineering (CE) case in Table 8-33 is selected as a bounding case for Westinghouse designed plants like Millstone, Unit 3, which is a conservative assumption. As discussed in Section 4.5.1 of EPRI Report 3002015906, Westinghouse and CE SG designs are similar, but the CE design is considered bounding for Westinghouse designs because the CE vessel is larger and has two primary outlet nozzles. A larger SG size will result in greater pressure stresses in the head, so the CE design inlet nozzle was chosen to represent both plant designs for B3.130 welds. Additionally, examination coverage increased in all four B3.130 welds in subsequent inspection intervals to at least 70 percent in all cases and to 100 percent for two of the welds in the fourth ISI interval. Based upon these factors, the NRC staff determined that the 55.5 percent examination coverage for these four B3.130 welds during the first and second ISI intervals would not lead to a probability of failure exceeding 1x10-6 failures per year if with examinations deferred for two 10-year intervals.
Based on this discussion, the NRC staff finds the Millstone, Unit 3, inspection history of the subject PZR and SG welds to be acceptable. Based on the above and the inspection history of the PZR and SG welds and NIR of Millstone, Unit 3, the NRC staff finds that the PFM approach of EPRI Reports 3002015905, 3002015906, and 3002014590 sufficiently represent the requested components for Millstone, Unit 3, with respect to ISI schedule and examination coverage.
4.9 Other Considerations The NRC staff reviewed the application concerning initial flaw depth and length distribution, probability of detection, models, uncertainty, and convergence. The NRC staff noted that these other considerations of the analyses in the EPRI reports do not depend on plant-specific information, as compared to component geometries, materials, and transient selection for which the licensee provided plant-specific information to ensure applicability of the analyses in the reports, as discussed previously.
Initial flaw depth and length distribution do not depend on plant-specific information because the flaw distribution used was based on fabrication flaws instead of service-induced flaws.
Probability of detection, which is associated with volumetric examinations, does not depend on plant-specific information because the corresponding components in different plants are subject to the same volumetric examination requirements of the ASME Code,Section XI. The models (i.e., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in fracture mechanics analyses. Uncertainty and convergence do not depend on plant-specific information because these are part of the overall PFM analyses that were addressed in the sensitivity studies and sensitivity analyses in the EPRI reports.
The NRC staff previously reviewed the applicable aspects of these considerations as used in EPRI Reports 3002015905, 3002015906, and 3002014590, and documented their acceptability in detail in the Salem SE, Millstone Unit 2 SE, and Vogtle SE. Since these considerations are no dependent on plant-specific information, the NRC staff finds that the plant-specific Millstone, Unit 3, submittal is acceptable in terms of these considerations.
4.10 PFM Results Relevant to Proposed Alternative In Section 5.0 of Attachment 1 its submittal, the licensee stated that based on the PFM results, after PSI, no other inspections are required for up to 80 years of plant operation to meet the acceptance criterion of 1x10-6 failures per year. Similar statements are made in EPRI Reports 3002015905, 3002015906, and 3002014590. 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.
Notwithstanding the discussion above, the PFM analyses in the EPRI reports investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. The PFM results relevant to the proposed alternative for Millstone, Unit 3, are those resulting from an ISI schedule scenario that closely matches that of Millstone, Unit 3, discussed in Section 4.8 of this SE. The relevant PFM results show that the probability of rupture is below the acceptance criterion of 1x10-6 failures per year. Based on the above and the discussions in sections 4.1 through 4.9 of this SE, the NRC staff finds that the proposed alternative for Millstone, Unit 3, for the requested PZR and SG welds and NIR of Millstone, Unit 3, would result in a PoF per year that is below the acceptance criterion of 1x10-6 failures per year.
4.11 Performance Monitoring 4.11.1 Background Performance monitoring, such as inservice inspection programs, is a necessary component such as described by the NRC five principles of risk-informed decision making. Analyses, such as PFM, work along with performance monitoring to provide a mutually supporting and diverse basis for facility condition and maintenance that is within its licensing basis. An adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation of continued appropriateness of associated analyses, and a timely method to detect novel/unexpected degradation. These characteristics were presented, for example, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277; agenda and slides, respectively). Previously, the NRC staff has applied binomial statistics and Monte Carlo methods to augment evaluation of periods beyond 20 years. The methods used by the NRC staff were presented at a May 25, 2022, public meeting (ML22144A345, and ML22143A840, meeting notice and presentation respectively).
4.11.2 Millstone, Unit 3, Evaluation The licensee provided the proposed examination schedule for the subject welds and NIR in Table 4 in Attachment 1 to the submittal. The proposed alternative for Millstone, Unit 3, would result in approximately 20-year spans between examinations for the welds in the submittal. The 20-year span is consistent with prior precedent where U.S. licensees have sought examination relief from prescriptive ASME Section XI requirements. In the submittal, the licensee also stated that the subject welds and NIR will be reexamined prior to the end of the current 60-year operating license for Millstone, Unit 3, which expires on November 25, 2045.
In its submittal, the licensee discussed system leakage tests as providing assurance of safety for the proposed alternative. However, 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 PZR and SG welds. The NRC staff noted that leakage tests provide complementary additional performance monitoring to the ISI examinations but would not, in isolation, be sufficient.
Prolonged periods without inspection may result in a lack of monitoring and trending capacity and provide weak basis for continued adequacy of component integrity. Consequently, the NRC staff performed a variety of simulations regarding potential inspection scenarios and the likelihood that such proposals would support the necessary characteristics of adequate performance monitoring. The NRC staff sought to understand the capacity of the proposed performance monitoring plan to detect potential novel degradation. These simulations were conducted using binomial statistics and Monte Carlo methods. Based on these simulations which encapsulates the Millstone, Unit 3, conditions, the NRC staff determined that the previously conducted and proposed volumetric inspections would constitute sufficient performance monitoring in concert with the other aspects of the submittal reviewed by the NRC staff. The NRC staff noted that some supporting monitoring and trending information will continue to be accrued at other facilities, spread by date of application, interval schedules, and other factors, providing further assurance that adequate monitoring and trending will continue.
Based on the above, 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. The NRC staff noted that the subject welds (and associated NIR where applicable) will be reexamined prior to the end of the current 60-year operating license for Millstone, Unit 3, which expires on November 25, 2045.
5.0 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 of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative for Millstone, Unit 3, to increase the ASME Code,Section XI, 10-year ISI interval for the subject welds (and associated NIR where applicable),
thereby deferring the examinations for those welds and NIR for two 10-year ISI intervals from the last examination performed for each item.
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: S. Levitus, NRR D. Dijamco, NRR Date: January 21, 2025
ML25014A286 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NAME TEdwards RGuzman KEntz DATE 1/13/2025 1/14/2025 1/15/2025 OFFICE NRR/DNRL/NVIB/BC (A)
NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME OYee HGonzález RGuzman DATE 11/29/2024 1/21/2025 1/21/2025