ML20352A155

Proposed Relief Request for Alternative VEGP-ISI-ALT-04-04 to the Requirements of the ASME Code

ML20352A155
Person / Time
Site:	Vogtle
Issue date:	01/11/2021
From:	Markley M Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Lamb J
References
EPID L-2020-LLR-0109
Download: ML20352A155 (30)

Text

January 11, 2021 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - RELIEF REQUEST FOR PROPOSED INSERVICE INSPECTION ALTERNATIVE VEGP-ISI-ALT-04-04 TO THE REQUIREMENTS OF THE ASME CODE (EPID L-2020-LLR-0109)

Dear Ms. Gayheart:

By letter dated December 11, 2019, as supplemented by letters dated September 9, 2020, and November 23, 2020, Southern Nuclear Operating Company (SNC, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the steam generator main steam outlet nozzle-to-vessel welds and steam generator feedwater nozzle-to-vessel welds and nozzle inside radius sections of the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),

SNC proposed to increase the ISI interval for the requested components to 30 years, from the current ASME Code Section,Section XI requirement of 10 years. The regulation in 10 CFR 50.55a(z)(1) requires SNC to demonstrate that the proposed alternative provides an acceptable level of quality and safety. SNC cited the analyses in Electric Power Research Institute (EPRI) Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections (Non-Proprietary), April 2019, in its proposed alternative. SNC evaluated the applicability of the EPRI report to Vogtle. The NRC staff reviewed the proposed relief request for Vogtle as a plant-specific alternative. The NRC did not review the EPRI report for generic use, and this approval does not extend beyond the Vogtle plant-specific authorization.

The NRC staff determined that SNCs proposed alternative in VEGP-ISI-ALT-04-04 to increase the ISI interval from 10 years to 30 years 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 in VEGP-ISI-ALT-04-04 for Vogtle for the duration of the fourth 10-year ISI interval through the sixth 10-year ISI interval.

All other requirements of ASME Code,Section XI, for which an alternative was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, contact John G. Lamb, Senior Project Manager, at 301-415-3100 or by e-mail to John.Lamb@nrc.gov.

Sincerely, Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425 cc: Listserv Michael T.

Markley Digitally signed by Michael T. Markley Date: 2021.01.11 09:15:13 -05'00'

Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR ALTERNATIVE VEGP-ISI-ALT-04-04, VERSION 1.0 STEAM GENERATOR MAIN STEAM AND FEEDWATER NOZZLE-TO-VESSEL WELDS AND NOZZLE INNER RADII ASME CODE, SECTION XI, EXAMINATION CATEGORY C-B SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

By letter dated December 11, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19347B105), as supplemented by letters dated September 9, 2020, and November 23, 2020 (ADAMS Accession Nos. ML20253A311 and ML20329A302, respectively), Southern Nuclear Operating Company (SNC, the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) a proposed alternative to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the steam generator (SG) main steam (MS) outlet nozzle-to-shell welds (NSWs) and SG feedwater (FW) NSWs and nozzle inside radius (NIR) sections of the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),

SNC proposed to increase the ISI interval for the requested components to 30 years, from the current ASME Code Section,Section XI requirement of 10 years. 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. SNC cited the analyses in a non-proprietary Electric Power Research Institute (EPRI) Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections (Non-Proprietary),

April 2019 (ADAMS Accession No. ML19347B107), in its proposed alternative. The licensee evaluated the applicability of the EPRI report to Vogtle. The NRC staff reviewed the proposed relief request for Vogtle as a plant-specific alternative.

The NRC did not review the EPRI report for generic use, and this relief request does not extend beyond the Vogtle plant-specific authorization.

2.0 REGULATORY EVALUATION

The SG MS outlet NSWs and SG FW NSWs and NIR sections at Vogtle are ASME Code Class 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, that are classified as ASME Code Class 1, 2, and 3 components shall 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 TECHNICAL EVALUATION

3.1 Components Affected Code Class:

ASME Code,Section XI, Class 2 Examination Category:

C-B, Pressure Retaining Nozzle Welds in Pressure Vessels Item Number:

C2.21 for the SG MS and FW nozzle-to-vessel welds C2.22 for the SG FW NIR sections Component Numbers:

As listed below Component ID Title 11201-B6-001-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-001-W19 16 MAIN FW NOZZLE TO SHELL WELD 11201-B6-002-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-002-W19 16 MAIN FW NOZZLE TO SHELL WELD 11201-B6-003-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-003-W19 16 MAIN FW NOZZLE TO SHELL WELD 11201-B6-004-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-004-W19 16 MAIN FW NOZZLE TO SHELL WELD 11201-86-001-IR04 MAIN FW NOZZLE INNER RADIUS 11201-B6-002-IR04 MAIN FW NOZZLE INNER RADIUS 11201-B6-003-IR04 MAIN FW NOZZLE INNER RADIUS 11201-B6-004-IR04 MAIN FW NOZZLE INNER RADIUS 21201-B6-001-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-001-W19 16 MAIN FW NOZZLE TO SHELL WELD 21201-B6-002-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-002-W19 16 MAIN FW NOZZLE TO SHELL WELD 21201-B6-003-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-003-W19 16 MAIN FW NOZZLE TO SHELL WELD 21201-B6-004-W18 32 STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-004-W19 16 MAIN FW NOZZLE TO SHELL WELD 21201-B6-001-IR04 MAIN FW NOZZLE INNER RADIUS 21201-B6-002-IR04 MAIN FW NOZZLE INNER RADIUS 21201-B6-003-IR04 MAIN FW NOZZLE INNER RADIUS 21201-B6-004-IR04 MAIN FW NOZZLE INNER RADIUS 3.2 Applicable ASME Code Edition Vogtle is currently in the fourth 10-year ISI interval (May 31, 2017, to May 30, 2027), and the ASME Code of record for this ISI interval is the 2007 Edition (2008 Addenda) of the ASME Code,Section XI.

Plant 10-Year ISI Interval ASME Code of Record Date

Vogtle, Units 1 and 2 4th 2007 Edition and 2008 Addenda 5/31/2017 to 5/30/2027

Vogtle, Units 1 and 2 5th 5/31/2027 to 5/30/2037

Vogtle, Units 1 and 2 6th 5/31/2037 to 5/30/2047 3.3 ASME Code Requirements Examination Categories, Examination Category C-B, Item No. C2.21, requires a surface and volumetric examination of the SG MS and FW NSWs at terminal ends of piping runs during each Section XI inspection interval. Examination Category C-B, Item No. C2.22 of Table IWC-2500-1 requires a volumetric examination of terminal ends of piping runs the SG FW NIR sections during each Section XI inspection interval. For both C2.21 and C2.22, the required volume (and surface for C2.21) are shown in the appropriate figure in IWC-2500-4(a), (b), or (d). As noted in Table IWC-2500-1 for Examination Category C-B, for cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.

3.4 Reason for Proposed Alternative In Section 5.0 of Enclosure 1 to its submittal dated December 11, 2019, SNC stated that EPRI performed an assessment of the basis for the ASME,Section XI examination requirements specified for Examination Category C-B, of ASME Section XI, Division 1 for SG, MS, and FW nozzle-to-shell welds and nozzle inside radius sections. The assessment includes a survey of inspection results from 74 units, as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The EPRI report concluded that the ASME Code Section XI interval of 10 years can be increased significantly with no impact to plant safety. SNC proposed to use this assessment to justify a proposed alternate ISI interval described in Section 3.5 of this SE.

3.5 Proposed Alternative In Section 6.0 of Enclosure 1 to its submittal dated December 11, 2019, the licensee stated that the proposed alternative is to increase the ISI interval for the requested components described in Section 3.1 of this safety evaluation (SE) from the current ASME Code,Section XI requirement of 10 years to 30 years.

3.6 Duration of Alternative The licensee requested to apply the alternative for the remainder of the fourth 10-year ISI interval and through the sixth 10-year ISI interval of Vogtle. The fourth 10-year ISI interval began on May 31, 2017, and the sixth 10-year ISI interval is currently scheduled to end on May 30, 2047.

3.7 Basis for Proposed Alternative In Section 6.0 of Enclosure 1 to its submittal dated December 11, 2019, SNC discussed the key aspects of the technical basis in the EPRI report and its applicability to Vogtle. These key aspects are summarized below.

Degradation Mechanisms Based on the evaluation of potential degradation mechanisms that could affect reliability of the SG MS and FW NSWs and NIRs (e.g., stress corrosion cracking, environmentally assisted fatigue (EAF), flow-accelerated corrosion, microbiologically influenced corrosion, pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated-corrosion, general corrosion, galvanic corrosion, and mechanical/thermal fatigue), the licensee stated that other than EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG MS and FW nozzles.

Stress Analysis The licensee stated that finite element analyses (FEA) performed in the EPRI report to determine the stresses in SG MS and FW NVWs and NIR sections based on representative PWR geometries, bounding transients, and typical material properties. In Appendix A, Vogtle Unit 1 and Unit 2 Applicablity, of Enclosure 1 to the supplement dated September 9, 2020, the licensee evaluated the applicability of the representative stress and flaw tolerance analyses in the EPRI report to the SG NVWs and NIR sections of Vogtle and determined that since the plant-specific requirements are met per Section 9.0 of the EPRI report, the results and conclusions of the report are met for applicability to Vogtle, Units 1 and 2.

Flaw Tolerance The licensee stated that flaw tolerance evaluations in the EPRI report consisted of PFM and DFM. SNC further stated that the results of the PFM analyses in the EPRI report indicate that, after a preservice inspection (PSI), no other inspections are required for up to 60 years of operation to meet the NRCs safety goal of 10-6 failures per year.

The licensee stated that Vogtle, Units 1 and 2, received PSI examinations followed by three 10-year interval inspections, and based on Table 8-10 in the EPRI report and if the ISI interval is increased to 30 years, the NRCs safety goal would be met with considerable margin for up to 80 years of plant operation.

The licensee stated that DFM analysis verified the PFM results by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to the ASME Code,Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (SIF) exceeds the ASME Code,Section XI allowable fracture toughness.

The licensee stated that DFM analysis verified the PFM results by demonstrating that a postulated flaw with an initial depth equal to the ASME Code,Section XI acceptance standards would take approximately 80 years to grow to a depth at which the maximum stress intensity factor (SIF) would exceed the fracture toughness specified in the ASME Code,Section XI. In the supplement dated September 9, 2020, SNC stated that the EPRI report was developed consistent with an EPRI White Paper on PFM and is noted in the supplement as Reference 14, EPRI Letter Number 2019-016 to NRC entitled, White Paper on Suggested Content for PFM Submittals to the NRC, dated February 27, 2019 (ADAMS Accession No. ML19241A545).

Inspection History In Appendix B of Enclosure 1 to the supplement dated September 9, 2020, the licensee discussed the inspection history to date of a subset of the requested components that were selected for examination (see Section 3.3 of this SE). The licensee showed in this appendix that the Vogtle SG FW and MS NSWs had limited coverage and that no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

Additionally, Structural Integrity Associates (SIA) Letter Report 1900064.406.R0, included as in the supplement dated September 9, 2020, stated that the requested components received an assumed 100 percent coverage during the preservice inspections, based on the combination of radiographic testing and the collective initial ASME Code,Section III and Section XI examinations, even though ultrasonic testing was not performed on the subject components.

The licensee also presented in Appendix C of Enclosure 1 to the supplement dated September 9, 2020 an industry-wide inspection history of components similar to the licensees requested components as supporting evidence that the components are flaw tolerant.

Conclusion The licensee concluded that the Vogtle SG MS and FW NSWs and NIR sections are very flaw tolerant. Having demonstrated the applicability of the analyses in the EPRI report to Vogtle, the licensee concluded that after PSI, no other inspection is required until 60 years to meet the NRCs safety goal of 1 x 10-6 failures per year and that an inspection interval of 30 years provides an acceptable level of quality and safety in lieu of the 10-year inspection interval required by ASME Code,Section XI, Examination Category C-B, Item Nos. C2.21 and C2.22.

The licensee concluded that the requested Vogtle components have performed with very high reliability as evidenced by the examination history of the components. The licensee noted that all other inspection activities, included the system leakage test (ASME Code,Section XI, Examination Category C-H) conducted each inspection period will continue to be performed, providing further assurance of safety.

3.8

NRC Staff Evaluation

The NRC staff reviewed SNCs technical basis for the proposed alternative pursuant to 10 CFR 50.55a(z)(1). The licensee referred to the results of the PFM analyses in the EPRI report as the primary basis for proposing to increase the ISI interval for the requested components from 10 years to 30 years. The NRC staffs review focused on evaluating the PFM analyses in Section 8.2 of the EPRI report and verifying whether the DFM analyses in Section 8.3 of the report support the PFM results. The NRC staff reviewed the proposed relief request for Vogtle as a plant-specific alternative. The NRC did not review the EPRI report for generic use, and this relief request does not extend beyond the Vogtle plant-specific authorization.

The degradation mechanism assumed in both the PFM and DFM analyses in the EPRI report that would lead to the failure criteria was fatigue crack growth (FCG). See discussion in Section 3.8.1, Overall PFM Approach, of this SE for a description of the failure criteria. The licensee referred to the evaluation of potential degradation mechanisms in Chapter 6 of the EPRI report and concluded that other than EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG MS and FW nozzles. EPRI did not analyze initiation due to mechanical/thermal fatigue or stress corrosion cracking in the PFM and DFM analyses in the EPRI report. Since fatigue initiation is one of the phenomena associated with EAF, the analyses in the EPRI report did not include EAF effects on initiation.

Based on the above, the NRC staff determined that the only degradation mechanism is cracking due to mechanical/thermal fatigue (i.e., FCG). EAF effects on FCG are included in the FCG rate selected for analyses (discussed in Section 3.8.6 of this SE). The NRC staff finds this acceptable for the plant-specific relief request since (1) assuming an initial crack is conservative because the time to initiate a crack would not factor into the analyses; (2) because FCG is known to be the dominant crack driving force in ferritic materials such as the SG MS and FW nozzles of Vogtle; and (3) ferritic materials are known to be highly resistant to stress corrosion cracking under the operating conditions of the requested SG nozzles.

3.8.1 Overall PFM Approach SIA developed the software analytical methodology PRobabilistic OptiMization of InSpEction (PROMISE) Version 1.0 to perform the PFM analyses in Section 8.2 of the EPRI report. On June 29, 2020; July 1, 2020; and July 27, 2020, the NRC staff conducted an audit of PROMISE Version 1.0 to verify that it properly implements PFM principles and has undergone adequate verification and validation (V&V). The NRC staffs audit plan was issued by letter dated May 14, 2020 (ADAMS Accession No. ML20128J311), and the audit report summary by letter dated December 10, 2020 (ADAMS Accession No. ML20258A002). The audit enabled the NRC staff to develop a better understanding how PFM principles were implemented in PROMISE Version 1.0 and the adequacy of V&V performed on the software. The audit focused on the information needs for a plant-specific authorization for Vogtle and did not review the EPRI report for generic use.

The overall PFM approach in the EPRI report is based on a Monte Carlo sampling technique in which PROMISE Version 1.0 samples parameters with statistical distributions (random parameters) many times to calculate a probability. Each sampling of parameters is known as a trial or a realization (see discussion on Uncertainty for the number of realizations used in the analysis).

As the NRC staff observed in Items 2.a.iv.1 and 2.a.iv.2 of the audit report, for each realization, PROMISE Version 1.0 performs a deterministic fracture mechanics analysis to calculate a time to failure to develop a histogram of failure times, which is, briefly stated, a tally of failure times.

Section 8.2.2.9 of the EPRI report defines failure as either rupture or leakage. Rupture is considered to occur when the applied stress intensity factor (SIF) exceeds plane strain crack initiation fracture toughness (KIC). Leakage occurs when the crack depth exceeds 80 percent of the wall thickness. The NRC staff noted that the PFM analyses is based on linear elastic fracture mechanics (LEFM).

From the histogram of failure times, PROMISE Version 1.0 estimates the probability of failure (PoF) at a given time as the fraction of the total number of realizations that the computed failure time is less than the given time. The PoF is then compared to the acceptance criterion of 1x10-6 per year.

Because the Monte Carlo technique is a widely used and accepted technique for calculating probabilities and counting times to failure is counting the number of failures (i.e., the probability that the failure time is less than a given time is equivalent to the probability that a failure would occur within that given time), NRC staff finds the overall PFM approach acceptable for the plant-specific Vogtle relief request.

The NRC staff noted that acceptance criterion of 1x10-6 per year 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. In that rule, the reactor vessel through-wall crack frequency of 1x10-6 per year for a pressurized thermal shock (PTS) event is acceptable. This is due to the limited impact of this failure frequency on the plant safety and its consistency with the NRC safety goals.

In addition, the NRC staff noted that acceptance criterion of 1x10-6 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 (ADAMS Accession No. 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. (ADAMS Accession No. ML19297G738).

Based on the above, the NRC staff finds the use of the acceptance criterion of 1x10-6 per year for the subject SG NSWs and NIR components acceptable for the plant-specific relief request.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.2 Parameters Most Significant to PFM Results In a PFM analysis, there are various input parameters that contribute to the final PoF value.

Examples include crack dimensions, fracture toughness, stress, crack growth rate, and ISI schedule, all of which may be further defined by sub-parameters (such as the exponent term in the crack growth rate). Analysts typically use two sensitivity tools to understand the effects of the input parameters. Sensitivity analyses (SA) help identify the major contributors to the final PoF value, and sensitivity studies (SS) help in determining the impact of each in parameter to the final PoF value.

In Section 8.2.4.2 of the EPRI report, EPRI performed SA to determine the dominant parameters that contribute to the probability of leak and ruptures. The results of these sensitivity analyses are in Tables 8-11 and 8-12 of the report for one of the locations analyzed.

For probability of leakage, EPRI determined that the most dominant contributor is FCG rate coefficient, and for probability of rupture, EPRI determined that the most dominant contributor is fracture toughness.

The NRC staff reviewed the overall results of the SA for the plant-specific relief request. Onset of leakage is driven by growth of the postulated crack by the FCG rate, which is a measure of how fast the postulated crack would grow to 80 percent of the wall thickness; and FCG rate is proportional to the FCG rate coefficient. Thus, the NRC staff finds that the FCG rate coefficient being the dominant contributor to probability of leakage reasonable. Rupture is driven by applied SIF (which is driven by stress) or fracture toughness since applied SIF and fracture toughness (represented by the parameter KIC) are the two parameters in the governing expression in LEFM: applied SIF < KIC. Thus, the NRC staff finds that fracture toughness being the dominant contributor to probability of rupture reasonable. The NRC staff noted that even though applied SIF did not come out as the dominant contributor in the SA, it is one of the significant parameters (reflected in the parameter of stress) in the SS, as discussed in the next paragraph.

In Section 8.2.4.3 of the EPRI report, EPRI performed SS on the following parameters: stress, fracture toughness, initial crack depth, number of flaws, flaw density (in the NIR), crack size distribution, FCG rate, probability of detection (POD), ISI schedule, and number of realizations.

EPRI concluded that the most significant parameters are stress, fracture toughness, and flaw density in the NIR. As with the SA results for probability of rupture, the NRC staff finds this overall result of the SS reasonable for the plant-specific relief request since these are the parameters that directly affect the governing expression in LEFM (flaw density in the NIR affects the applied SIF side of the expression).

During the audit of PROMISE Version 1.0, the NRC staff observed that ISI schedule and examination coverage have a significant impact on the PoF. The NRC staff requested two SIA letter reports that cover these topics, 1900064.406.R0 and 1900064.407.R2, which were included as Enclosures 2 and 3, respectively, in the supplement dated September 9, 2020.

The SA, SS, and the NRC staffs observations during the audit of PROMISE Version 1.0 on the effects of ISI schedule and examination coverage thus identified the following significant parameters or aspects of the PFM analyses that warrant a close evaluation: stress analysis, fracture toughness, flaw density, 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 3.8.8, Other Considerations, of this SE.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.3 Stress Analysis 3.8.3.1 Selection of Components and Materials The licensee evaluated the plant-specific applicability of the components and materials selected in the EPRI report to the SG MS and FW nozzles of Vogtle (Appendix A of Enclosure 1 to the September 9, 2020, supplement).

The licensee showed that Vogtle meets the component configuration and material criteria. The acceptability of meeting the criteria, however, depends on the acceptability of the component and material selection described in the EPRI report, which the NRC staff evaluated below.

In Section 4.5 of the EPRI report, EPRI discussed the selection of three representative SG nozzle components - two MS nozzles and one FW nozzle, for stress analysis. 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.

EPRI concluded that variations in the design of SG MS and FW nozzle among the three main nuclear steam supply system vendors (Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W) are not significant, and that the most important parameter, ratio of radius-to-thickness (R/t) ratio (of the SG shell or MS and FW piping attached to the requested nozzles) can be addressed by sensitivity studies in the PFM evaluation. Figure 4-4 of the EPRI report shows the various R/t ratios considered and identifies the ones selected for analysis. In Section 4.6 of the report, EPRI stated that it did not address the auxiliary FW nozzles and CE System 80 MS and FW nozzle designs.

The NRC staff reviewed Sections 4.1 through 4.6 of the EPRI report and finds the three SG nozzles selected for stress analysis as acceptable representatives of SG MS and FW nozzles for the Vogtle plant-specific relief request, because differences in R/t ratios are small, and therefore, differences in stresses would be small and can be reasonably addressed through sensitivity studies. The NRC staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations is acceptable for the Vogtle plant-specific relief request since R/t ratio is related to the pressure stress, and pressure stress is the dominant stress as evidenced by the through-wall stress distributions in Sections 7.1 through 7.3 of the EPRI report.

The NRC staff noted from Figure 4-4 of the EPRI report that the Westinghouse 4-loop and Babcock & Wilcox MS nozzle selected for analysis is adequately representative of their respective configurations, as the R/t variations are within 10 percent. The NRC staff noted that there is only one B&W configuration in Figure 4-4. The NRC staff also reviewed the evaluation in Section 7.4 of the report that addresses the other two Westinghouse SG MS nozzle configurations - the reduced reinforcement and tapered designs. From Figure 7-41 of the EPRI report, the NRC staff noted that at the NIR, the hoop pressure stress ratio could be up to 1.5, which means the hoop stress could be about 1.5 times higher relative to the Westinghouse SG MS nozzle selected for analysis. At the NSW, Figure 7-42 shows the differences of meridional stress from the Westinghouse SG MS nozzle selected for analysis. The NRC staff noted that in addition to the meridional stress, the hoop stress comparison needs to be shown in the NSW since axial flaws (driven by hoop stress) were also evaluated in the PFM analysis at the NSW.

Figures 7-37 and 7-39 of the EPRI report show the hoop stress contour plots for the tapered and reduced reinforcement designs, respectively. The NRC staff compared these contour plots to the hoop stress plot in Figure 7-5 for the Westinghouse SG MS nozzle selected for analysis.

From this comparison, the NRC staff determined independently that at the NSW, the hoop stress could be up to over 2.2 times higher relative to the Westinghouse SG MS nozzle selected for analysis.

For the Westinghouse FW nozzle, the NRC staff noted from Figure 4-4 of the EPRI report that where the SG shell R/t ratio (x-axis of Figure 4-4) was larger than the FW nozzle selected for analysis, the variation is within 10 percent. However, the FW nozzle bore R/t ratio (y-axis of Figure 4-4) can be up to 1.85, while for the selected FW nozzle configuration, the nozzle bore R/t is only about 1.38. Thus, the pressure stress at the bore for other FW nozzles could be up to 1.85/1.38 = 1.34 times higher.

In Section 8.2.4.3.2 of the EPRI report, EPRI performed sensitivity studies (SS) on the impact of stress by determining what multipliers to the base case stresses (those determined from the stress analysis) would lead to exceedance of the PoF criterion of 1x10-6 per year. The result of the study showed that a stress multiplier of 2.2 would lead to exceedance of the criterion for the MS nozzles and 1.5 for the FW nozzle. The NRC staff noted that the multiplier of 2.2 adequately addresses the variation of up to 2.2 times the hoop pressure stress discussed above for the Westinghouse SG MS nozzle. The NRC staffs estimate of up to 2.2 times was a conservative estimate based on comparing stress contour plots that showed stress ranges. A more realistic estimate is about 1.7 times. For the FW nozzle, the NRC staff determined that the multiplier 1.5 times in the sensitivity study adequately addresses the variation of 1.34 times the hoop pressure stress at the nozzle bore.

The NRC staff noted that, as discussed in the EPRI report, the EPRI report does not apply to the auxiliary FW nozzles and Combustion Engineering System 80 MS and FW nozzle designs, since EPRI did not address these nozzles in the report.

In Section 5 of the EPRI report, EPRI discussed the material properties for SA-533 Grade A Class 2 and SA-516 Grade 70 that were selected for the stress analysis. The NRC staff finds these two materials acceptable for the Vogtle plant-specific relief request because they are common ferritic materials used in SG vessels as discussed in Section 4.3 of the EPRI report.

In Appendix A of Enclosure 1 to its supplement dated September 9, 2020, the licensee shows that Vogtle meets the component configuration and material criteria in the EPRI report.

Therefore, the NRC staff finds that the component configuration and materials of the requested SG nozzles of Vogtle are acceptable.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.3.2 Selection of Transients The licensee evaluated the Vogtle plant-specific applicability of the transients selected in the EPRI report to the SG MS and FW nozzles of Vogtle (Appendix A of Enclosure 1 to the submittal). The licensee stated that the cycles in Table 5-5 of the EPRI report meet or exceed the 60-year projected cycles for Vogtle that were based on fatigue monitoring data.

The acceptability of meeting the transient criterion, however, depends on the acceptability of the transient selection described in the EPRI report, which the NRC staff evaluated below.

In Section 5.2 of the EPRI report, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to the SG MS and FW nozzles. EPRI showed a comparison of design cycles with projected cycles based on transient monitoring systems and explained that even though design cycles for loading and unloading transients are in the range of 20,000, the projected cycles for these transients are less than 1,000 for 60 years of operation because many plants do not load-follow.

In Table 5-4 of the report, EPRI showed a sample list of design transients for each of the three nuclear steam supply system reactor types (Westinghouse, CE, and B&W). EPRI developed a list of transients for analysis, shown in Table 5-5 of the EPRI report, based on transients that have the largest temperature and pressure variations. Table 5-5 shows the temperatures at the steam space (TST) and those below the steam space (TDC and TFW). Transients that occur in the steam space affect the MS nozzle, and those that occur below the SG steam space affect the FW nozzle. Table 5-5 also shows minimum and maximum pressures.

EPRI stated additional cycles of the loss-of-load transient addressed transients not explicitly selected for analysis. The NRC staff reviewed the discussion of cycles in Section 5.2; compared the Table 5-5 transients with those in the sample list of transients in Table 5-4, and considering that the transient selection was based on the largest temperature and pressure variations, which is conducive to FCG; determined that the transients defined in Table 5-5 of the EPRI report selected for analysis are reasonable for the Vogtle plant-specific relief request.

EPRI did not consider test conditions beyond a system leakage test. EPRI stated that since any pressure tests will be performed at operating pressure, no separate test conditions need to be included in the evaluation. The NRC staff reviewed the Vogtle Updated Final Safety Analysis Report (UFSAR) and noted that Sections 3.9.N.1.1.1.15 and 3.9.N.1.1.5.2 (ADAMS Accession No. ML19296C722) specify a secondary side leakage test performed at just below design pressure 80 times during plant life and a secondary side hydrostatic test performed at 1.25 of design pressure 10 times during plant life.

In terms of cycles, the NRC staff determined for the Vogtle plant-specific relief request that the heat-up and cooldown transient in Table 5-5 of the EPRI report reasonably accounts for the total of 90 cycles of pressure testing (system leakage and hydrostatic testing) of the secondary side of Vogtle. Table 5.4.2-1 of the Vogtle UFSAR (ADAMS Accession No. ML19296C741) specifies a design pressure of 1,185 pounds per square inch gauge (psig) for the secondary side. Thus, the NRC staff noted that the pressure during pressure testing (hydrostatic test) could be up to 1.25 x 1,185 = 1,481 psig. The NRC staff determined that the SS on stress reasonably address this higher pressure.

The minimum temperature specified for the system leakage and secondary side hydrostatic tests described above is 120 degrees Fahrenheit (°F). The NRC staff noted this minimum temperature is too low for the assumption of a KIC value of 200 ksiin to be appropriate (see Section 3.8.4 of this SE). The NRC staff further noted that the T - RTNDT value used in calculating the ASME Code KIC value must be at least 105 °F for the material to be on the upper shelf (KIC value of 200 ksiin). In Section 8.2.2.7 of the EPRI report, EPRI stated that it assumed an RTNDT value of 60 °F for the requested SG nozzles. Therefore, the temperature T in T - RTNDT must be at least 105 °F + 60 °F = 165 °F.

In order for the KIC value of 200 ksiin to be acceptable for Vogtle, the NRC staff needed confirmation that when the secondary side system leakage and hydrostatic tests at Vogtle are performed at the maximum pressures specified for the tests, the temperature is least 165 °F. In its supplement dated November 23, 2020, the licensee confirmed that the temperature during the secondary side system leakage tests at Vogtle is performed well above 165 °F, and, therefore, the assumed KIC value of 200 ksiin in the EPRI report is appropriate. The licensee also explained that the secondary side hydrostatic test is not applicable to Vogtle since the secondary side hydrostatic test was performed during construction, and, that if there are any repairs, the licensee opts for performing system leakage tests rather than hydrostatic tests.

The licensee further pointed to the results of the SS on KIC in Tables 8-13 and 8-14 of the EPRI report to show that a KIC value of 100 ksiin (appropriate for a test temperature of 120 °F given the assumed RTNDT value of 60 °F in the EPRI report) results in a probability of rupture below 1x10-6 per year. The NRC staff finds the licensees discussion on the appropriateness of the assumed KIC value of 200 ksiin during the secondary side system leakage and hydrostatic tests acceptable since the secondary side leakage tests are performed at temperatures well into the upper shelf temperatures at which a KIC value of 200 ksiin is valid and because hydrostatic tests are not applicable to Vogtle. The NRC staff also noted its evaluation of fracture toughness in Section 3.8.4 of this SE in which a KIC value as low as 80 ksiin produced rupture probabilities below 1x10-6 per year.

Based on the above, the NRC staff finds that the assumption of a KIC value of 200 ksiin is appropriate for the secondary side system leakage and hydrostatic tests at Vogtle, and therefore, test conditions at Vogtle are captured adequately by the other transients analyzed in the EPRI report.

EPRI also did not consider emergency and faulted conditions. In Section 5.2 of the EPRI report, EPRI assumed that during faulted condition transients, the temperature and pressure in the SG secondary side would decrease and that the decrease in temperature is expected to be small.

EPRI further stated that the drop in pressure would unload the system and reduce stresses in the requested components, and thus, the normal and upset condition transients selected for analysis, which include large pressure and temperature decreases (e.g., cooldown), bound the faulted condition transients. The NRC staff determined this to be a reasonable conclusion because the thermal transient stress would be small due to the small temperature decrease during these faulted events and because the pressure would be decreasing at the same time.

Also, considering that a faulted (or emergency) condition transient is rare (typically only one event during the life of a plant), the NRC staff determined that not including a faulted condition transient in the analysis is acceptable for the plant-specific relief request.

Note 6 in Table 5-5 of the EPRI report states that the loss-of-power transient affects only the FW nozzle, and therefore, was applied only to the FW nozzle analysis. The NRC staff noted that the output files for the limiting FW nozzle case, FEW-P3A.rpt and FEW-P3A-ISI.rpt, contain all the transients listed in Table 5-5 except for the loss-of-power transient (Items 2.e.i and 2.e.ii of the audit report). The through-wall stress distribution plots in Figures 7-32 through 7-35 of the EPRI report for the FW nozzle show a stress distribution for the loss-of-power transient at 619 seconds. The NRC staff also noted in Items 2.e.i and 2.e.ii of the audit report an LTOP event in the input files for the limiting FW nozzle case, FEW-P3A.vnz3 and FEW-P3A-ISI.vnz3.

The NRC staff noted that the EPRI report did not contain information on an low-temperature overpressure protection (LTOP) (which the NRC staff presumes to stand for low temperature overpressure) event, and that the design bases in Section 5.2.2.1 of the Vogtle UFSAR stated that the overpressure protection for the steam system is provide by the SG safety valves.

Therefore, it is not clear to the NRC staff the relevance of an LTOP event for the requested SG nozzles. The NRC staff needed explanations and/or clarifications on the loss-of-power transient and LTOP. In its supplement dated the November 23, 2020, the licensee explained that the cycles of the loss-of-load transient were increased from 100 to 360 to account for other transients, including the loss-of-power transient.

The licensee further demonstrated that the loss-of-power transient has insignificant impact on the PFM results by showing results with and without the loss-of-power transient. Thus, the NRC staff finds the licensees response regarding the loss-of-power transient acceptable for the Vogtle plant-specific relief request, because the licensee clarified that the transient was included in the PFM analyses and that its impact on the PFM results is insignificant. Regarding LTOP, the licensee clarified that LTOP events are applicable to the primary side and that the SG, a secondary side component, is protected from LTOP by SG safety relief valves. The NRC staff finds the licensees clarification regarding LTOP acceptable for the Vogtle plant-specific relief request, and determined that LTOP need not be included in the PFM analyses.

In reviewing the input files described above, the NRC staff also observed that the number of cycles in Table 5-5 of the EPRI report were converted into cycles per year. Thus, the NRC staff noted that even though the cycles in Table 5-5 are for a 60-year period, the crack growth analyses performed in the PFM analysis due to the listed transients are based on the Table 5-5 cycles that are converted into cycles per year (e.g., 5,000 cycles in 60 years is equivalent to 84 cycles per year) and the crack growth computed for up to 80 years.

The licensee showed in Table A2 of Appendix A of Enclosure 1 to its supplement the September 9, 2020, that the number of cycles of the selected transients for Vogtle is bounded by the number of cycles in the EPRI report. Therefore, the NRC staff finds that the transient loads for the requested SG nozzles of Vogtle are acceptable.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.3.3 Other Operating Loads In Section 5.2 of the EPRI report, EPRI stated that loads from the piping attached to the MS or FW nozzles are not considered in the analysis. The NRC staff finds this assumption acceptable for the Vogtle plant-specific relief request since bending stresses due to the attached piping have marginal impact on the orientation of the flaws postulated in the NSW and NIR.

In Section 8.2.2.4.2 of the EPRI report, EPRI discussed the cosine distribution assumed for the through-wall residual stress due to welding. EPRI treated this distribution as a non-random parameter and stated that since, in some configurations, the NSW is close to the NIR, this cosine distribution is applicable to both locations. EPRI stated that this residual stress distribution has been used in past projects, particularly BWRVIP-108 for which that the NRC staff has issued an SE dated December 19, 2007 (ADAMS Accession No. ML073600374).

BWRVIP-108 is the PFM-based technical basis for the reduction of the number of nozzles inspected in boiling-water reactor pressure vessels on which some of the inputs for the current EPRI report were based. The NRC staff noted that the residual stress distribution in BWRVIP-108 is for welding in thick-walled vessels. The NRC staff finds the cosine distribution EPRI assumed for the through-wall residual stress due to welding acceptable for the Vogtle plant-specific relief request since the requested SG NSWs and NIR sections are in SGs, which are thick-walled vessels.

The NRC staff noted from Figure A1 of Appendix A to Enclosure 1 of its supplement dated September 9, 2020, that the NIR location of the Vogtle SG FW nozzle is relatively far away from the NSW, and thus, the assumption of residual stress in the NIR for Vogtle is conservative. The NRC staff noted that since the SG MS and FW nozzles are not cladded, no cladding stress is needed. Thus, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested SG nozzles of Vogtle.

3.8.3.4 Finite Element Analyses In Section 7 of the EPRI report, EPRI discussed the FEA to determine stresses due to internal pressure and thermal transients for the three selected geometries discussed in Section 3.8.3.1 of this SE. The NRC staff reviewed the modeling details (elements used, boundary conditions, symmetry assumptions, etc.) and finds that they are consistent with standard FEA practice for the plant-specific relief request.

The NRC staff also reviewed the stress contour plots and the through-thickness stress distribution in the section and find them acceptable for the plant-specific relief request. For instance, the NRC staff noted the average hoop stress due to a 1,000 pounds per square inch (psi) internal pressure at the SG FW nozzle shell in Figure 7-34 is approximately 28,000 psi.

The NRC staff calculated a hoop stress of 1,000(R/t) = 1,000(25) = 25,000 psi (R/t = 25 for the SG FW nozzle shell from page 4-18 of the EPRI report).

The NRC staff noted that EPRI modeled the effect of the thermal sleeve in the SG FW nozzle by applying an effective heat transfer coefficient to the nozzle surface protected by the thermal sleeve (see Sections 7.3.1 and 7.3.2.2 of the EPRI report). The NRC staff also noted that EPRI aptly included in the plant-specific applicability criteria in Section 9 of the report that the FW nozzle design must have an integrally attached thermal sleeve. The licensee stated in Appendix A of Enclosure 1 to its supplement dated September 9, 2020, that the Vogtle FW nozzles have an integrally attached thermal sleeve.

Based on the above, the NRC staff determined that the thermal stresses calculated through FEA for the SG MS and FW nozzles apply to the requested SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use..

3.8.4 Fracture Toughness In Section 8.2.2.7 of the EPRI report, 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. As discussed in Section 3.8.3.1 of this SE, Vogtle meets the material criteria in the EPRI report, and thus the NRC staff determined that the fracture toughness parameters above are applicable to Vogtle.

The NRC staff had accepted the treatment of upper shelf KIC as a random parameter in BWRVIP-108 through sensitivity studies on KIC that the NRC staff requested, which changed the standard deviation to less than 5 ksiin and greater than 5 ksiin (see December 19, 2007, SE of BWRVIP-108; the EPRI report also includes SS on KIC as discussed below). Apart from this, EPRI stated in Section 8.2.2.7 of the EPRI report that the RTNDT assumed for the requested SG nozzles is 60 °F, which the NRC staff noted would have some conservative effect in the assumption of the upper shelf KIC value. For these reasons, the NRC staff finds that the mean and standard deviation values of upper shelf KIC used in the EPRI report are reasonable for the Vogtle plant-specific relief request, even though statistical distributions that would more accurately account for the uncertainty in upper shelf fracture toughness should have been used.

EPRI further explained that an upper shelf KIC value of 200 ksiin can be used for fracture toughness since the minimum temperature from all applicable transients is sufficiently high (i.e.,

the temperature in the requested components stays in the range in which the upper shelf value of 200 ksiin is applicable). The NRC staff discussed in detail with the licensee the transient temperature aspect of the assumption of 200 ksiin during the audit of PROMISE Version 1.0.

As the NRC staff observed during the audit (Item 2.e.iii of the audit report), some of the temperatures in the list of transients in Table 5-5 of the EPRI report are low enough to invalidate the assumption of 200 ksiin.

In response to the NRC staffs request, the licensee showed history plots of total applied SIF (with KIC) and temperature at the FW nozzle NSW and NIR for heat-up/cooldown, plant loading, and plant unloading transients. The NRC staff determined that showing only plots for the FW nozzle is sufficient since the heat transfer coefficient assumed in the FW nozzle was higher than for the MS nozzle. The higher heat transfer coefficient would result in higher thermal stresses in the FW nozzle. The NRC staff observed from the plots that total applied SIF can exceed KIC at a value as low as 130 ksiin at the FW nozzle NSW during the beginning and ending of the heat-up/cooldown transient.

The licensee explained that the results of the sensitivity studies on KIC in Tables 8-13 and 8-14 of the EPRI report address this low KIC value. Table 8-13 shows the probability of rupture for KIC

= 80 ksiin is 3.75x10-8 per year for the FW nozzle at the NSW (case FEW-P3A) and 2.50x10-9 per year for the limiting MS nozzle (case SGB-P3A), which are below the criterion of 1x10-6 per year. See Section 3.8.9 of this SE for discussion of the combined effects of stress, fracture toughness, and flaw density. The NRC staff confirmed during the audit the values in case FEW-P3A with those in the output file for that case. Also, considering that the total applied SIF showed in the plot history was for the deepest (and longest) flaw, the NRC staff determined that during the Monte Carlo sampling, the occurrence of this deepest flaw would be rare, given the statistical distribution on initial flaw depth (see Section 3.8.8.1 of this SE).

While the NRC staff finds for this Vogtle plant-specific relief request that the sensitivity studies in which KIC dropped to a value of 80 ksiin would reasonably account for the low temperatures during the beginning and ending of the heat-up/cooldown transient, modeling KIC as a function of temperature would have been a better approach, as there may be cases at the very beginning and very ending of heat-up/cooldown when KIC could drop even below 80 ksiin. This could lead to slightly higher PoF values, but the NRC staff determined for this Vogtle plant-specific relief request that the impact on PoF values would be minimal since pressure would be very low at the beginning and ending of heat-up/cooldown. The NRC has neither reviewed nor approved the EPRI report for generic use.

For the plant loading, plant unloading, and loss-of-power transients, the NRC staff observed that the temperature at the NVW and NIR of the FW nozzle remained high, approximately 500 °F or greater, even though the water temperature at the FW nozzle ranged from 32 °F to 430 °F.

Therefore, the NRC staff determined for the Vogtle plant-specific relief request that a KIC value of 200 ksiin is applicable. Even though the NRC staff made no observations about the loss-of-load transient, the NRC staff determined for the Vogtle plant-specific relief request that the temperature conditions at the NVW and NIR of the FW nozzle would be similar to the plant loading and plant unloading transients, because the temperatures for the three transients shown in Table 5-5 of the EPRI report are similar.

Based on the above and the discussion in Sections 3.8.3.1 and 3.8.3.2 of this SE, that the materials and transient loads of the requested SG nozzles of Vogtle are acceptable, the NRC staff finds the fracture toughness model acceptable for the requested SG nozzles at Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.5 Flaw Density In Section 8.2.2.2 of the EPRI report, EPRI stated that 1.0 fabrication flaw per nozzle is used at the NSW and 0.001 flaw at the NIR, and that these values are consistent with those the NRC staff approved in BWRVIP-108. As discussed in the SE of BWRVIP-108 dated December 19, 2007, the NRC staff had requested that the number of flaws per nozzle at the NIR be revised to 0.1 flaw per nozzle in the PFM analyses in BWRVIP-108, and therefore, the acceptable number of flaws at the NIR is 0.1 flaw per nozzle (discussed further three paragraphs down).

The NRC staff noted that the thicknesses at the requested locations analyzed in the EPRI report

- SG MS NSW and SG FW NSW and NIR, are generally less than the thickness of PWR vessels. Since the flaw density of the Pressure Vessel Research Users Facility (PVRUF) depth distribution was derived from fabrication flaw data from a PWR vessel (see discussion in Section 3.8.8.1 of this SE), the NRC staff evaluated the applicability of the assumed of flaws of 1.0 per nozzle at the NSW (and consequently of the assumed number of flaws at the NIR) to the thicknesses of the requested locations in the SG vessel.

The NRC staff determined in the December 19, 2007, SE of BWRVIP-108 that 1.0 flaw per nozzle is conservative in the NSW based on the surface-breaking flaw density derived from the PVRUF project and weld area of a representative BWR NSW. Since the Vogtle SG vessel thicknesses (shown in the figures in Appendix A of Enclosure 1 to the September 9, 2020 supplement) are only slightly thinner than BWR vessel thicknesses, and the NRC staff determined 1.0 flaw per nozzle was conservative for BWR NSWs, the NRC staff finds for the Vogtle plant-specific relief request that applying 1.0 flaw per nozzle for the SG vessel NSW is also conservative.

In Sections 8.2.4.3.4 and 8.2.4.3.5 of the EPRI report, EPRI performed SS on the number of flaws in the NIR and NSW to determine the impact of these parameters on the PoF results.

In Section 8.2.4.3.4 of the EPRI report, EPRI changed the number of flaws in the NIR from 0.001 to 0.1, and as a result, the probabilities of leak and rupture increased by two orders of magnitude, but were still significantly below the acceptance criterion of 1x10-6 per year. This result, however, is only the effect of changing to 0.1 flaw per nozzle at the NIR. Other significant parameters such as stress and fracture toughness must be considered. The NRC staff discussed the combined effects of stress, fracture toughness, and flaw density at the NIR in Section 3.8.9 of this SE. The NRC staff observed in Item 3.d of the audit report of PROMISE Version 1.0 that the flaw density value in the NIR is a multiplier on the probability values in the PROMISE Version 1.0 output. The NRC staff finds for the Vogtle plant-specific relief request that for the probability output values in the NIR, multiplying by the fractional flaw density is a reasonable approach to account for the lower number of flaws expected in the nozzle forging.

In Section 8.2.4.3.5, EPRI changed the constant, 1.0 flaw per nozzle in the NSW, to a Poisson distribution in which some realizations had more than one flaw, while other realizations may not have had any flaws, as the NRC staff observed in Item 2.b.ii of the audit report. The result was that the probability of leakage for the 1.0 flaw per NSW was only slightly higher than the Poisson distribution. Also, as observed in Item 2.b.ii of the audit report, the NRC staff noted that in cases where there was more than one flaw during a realization, the flaws did not interact.

The NRC staff finds for the Vogtle plant-specific relief request that even though no flaw interaction is slightly nonconservative, the impact of this nonconservatism to the final PoF results would be minimal.

Based on the above, the NRC staff finds for the Vogtle plant-specific relief request that the assumption of 1.0 flaw per nozzle at the NSW is acceptable for the corresponding NSWs of the requested SG nozzles of Vogtle. On the assumed number of flaws per nozzle at the NIR, the NRC staff finds for the Vogtle plant-specific relief request that the acceptable number of flaws is 0.1 per nozzle. Even though EPRI assumed 0.001 flaw per nozzle at the NIR for the base case PFM analyses, EPRI performed a SS on the impact of 0.1 flaw per nozzle. Considering this SS and the NRC staffs discussion of the results in Tables 8-15 and 8-16 of the EPRI report in Section 3.8.9 of this SE, the NRC staff determined that the EPRI report has adequate information for the Vogtle plant-specific relief request that addresses the effect of 0.1 flaw per nozzle at the NIR, compared to 0.001 flaw per nozzle for the corresponding NIR sections of the requested SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report generic use.

3.8.6 FCG Rate In Section 8.2.2.6 of the EPRI report, EPRI stated that the FCG rate for ferritic steels, as defined in the 2017 Edition of the ASME Code,Section XI, Appendix A, paragraph A-4300, is used in the evaluation. The NRC staff verified that the 2017 Edition of the ASME Code,Section XI, is the latest edition incorporated by reference in 10 CFR 50.55a.

Specifically, the FCG rate is defined with a log-normal distribution with the median value defined as the FCG rate in ASME Code,Section XI, A-4300, and with a value of 0.467 for the uncertainty parameter. The NRC staff finds the uncertainty parameter of 0.467 acceptable for the Vogtle plant-specific relief request because it is based on over 1,000 FCG rate data for low alloy steels.

In Section 8.2.4.1 of the EPRI report, EPRI stated that assuming the A-4300 curve as median is conservative since the actual data from which the A-4300 curve is based on represent the 95-percent confidence limit of the data. The NRC staff clarifies that 95-percent confidence limit here means that the A-4300 curve bounds the median of the data 95 percent of the time; it does not mean that the A-4300 curve is the 95th percentile of the data. Because of the amount of available data for ferritic FCG rate, however, the difference between 50-percent confidence limit on the median and 95-percent confidence limit on the median would likely be small. Thus, the NRC staff determined for the Vogtle plant-specific relief request that assuming the A-4300 as the median curve would only be slightly conservative.

EPRI stated that the associated threshold on the FCG rate is also log-normally distributed and that the log-normal distributions on the rate and threshold are consistent with the approach used in the xLPR, a PFM software sponsored by the NRC and EPRI. The NRC staff noted that the FCG rate is the rate defined in A-4300 with a statistical distribution around it since the FCG rate is treated as a random parameter. In Section 8.2.4.3.7 of the EPRI report, EPRI performed a SS on the effect of FCG rate on probability of leakage by replacing the A-4300 FCG rate with the FCG rate used in BWRVIP-108. The result of the study showed that the A-4300 FCG rate led to a much higher probability of leakage. The NRC staff noted that the FCG rate in ASME Code,Section XI, A-4300 is applicable to both BWR and PWR.

Therefore, the NRC staff finds for the Vogtle plant-specific relief request that the A-4300 FCG rate used in the analyses is acceptable for the results applicable to the requested SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.7 ISI Schedule and Examination Coverage In Section 8.2.4.1.2 of the EPRI report, EPRI discussed the effect of various ISI schedules on the PoF results. In Enclosure 2 to the submittal (SIA Report No.1900064.406.R0), the licensee addressed the effect of examination coverage. Examination coverage refers to the volume of the NSW or NIR that received examination during the ISI. In Section 6.0 of Enclosure 1 to the submittal, the licensee stated that the SG requested components of Vogtle received PSI examinations and three 10-year ISI examinations (PSI + 10 + 20 + 30). The licensee is proposing an alternative that would extend the ISI interval to 30 years after the first three 10-year examinations (i.e., an alternative ISI schedule of PSI + 10 + 20 + 30 + 60).

The NRC staff noted in Figure 8-9 of the EPRI report that the ISI schedule has a significant impact on the PoF values. The NRC staff documented in its audit report (Item 2.c.i) how PROMISE Version 1.0 implements ISI, as described in the following: the number and frequency of ISI are input into the software; at the specified times of ISI, flaws are either detected or not detected with the chance of detection/non-detection given by the POD curve (see Section 3.8.8.2 of this SE for further discussion of the POD curve). If detected, a flaw is assumed to be repaired or properly dispositioned, and thus, cannot cause failure; if not detected, the flaw continues to grow, and thus, can lead to failure. The NRC staff finds this to be a better approach than applying an adjustment factor to the failure probabilities since the effect of the POD curve would be propagated into the failure probabilities each time ISI is implemented. The NRC staff finds for the Vogtle plant-specific relief request that the approach described a reasonable approach for implementing the effect of ISI.

The NRC staff reviewed Enclosure 3 to the licensee supplement dated September 9, 2020 (SIA Report No. 1900064.407.R2), which shows the benchmarking of the ISI implementation of PROMISE Version 1.0, with another one of SIAs PFM software, VIPERNOZ. The comparison plots in Figures 1 through 4 of SIA Report No. 1900064.407.R2 showed adequate agreement between PROMISE Version 1.0 and VIPERNOZ. Even though the NRC has not formally accepted VIPERNOZ, it is the PFM software used in the BWRVIP-108 report for which the NRC staff has issued an SE. Thus, the NRC staff finds for the Vogtle plant-specific relief request that VIPERNOZ is a reasonable basis for the benchmark comparisons. Accordingly, the NRC staff determined for the Vogtle plant-specific relief request that the ISI algorithm in PROMISE Version 1.0 produces acceptable results.

The NRC has neither reviewed nor approved the EPRI report and VIPERNOZ for generic use.

In Section 8.2.5 of the EPRI report, EPRI stated that it assumed 100-percent inspection of the required volume (i.e., 100-percent examination coverage) of the requested SG nozzles during each of the ISI scenarios evaluated in the report, which assumes 100-percent examination coverage during PSI. EPRI further explained that based on its statement on the PSI-only examinations in Section 8.2.4.1.1 of the report, by performing complete 100-percent examination coverage during PSI, no other examinations are needed for safe plant operation for 80 years, and that based on this, any additional ISI examinations after PSI would reduce the already low PoF values (e.g., Table 8-9 of the EPRI report), and that, therefore, the PFM evaluations with 100-percent examination coverage assumed for all ISI also apply to partial (less than 100-percent) examination coverage.

In the first paragraph of Section 3.8.9 of this SE, the NRC staff explained its non-acceptance of the PSI-only examination conclusion. Thus, the NRC staff determined that partial examination coverage plays a vital role in the final PoF values. The NRC staff documented in the audit report (Item 2.c.i) how PROMISE Version 1.0 implements examination coverage for a case with 50-percent coverage. In a given PFM evaluation, the POD curve is not applied for approximately 50-percent of the number of realizations, and thus for 50-percent of the realizations, a postulated flaw would continue to grow. The NRC staff noted that a postulated flaw continuing to grow for 50-percent of the number of realizations could significantly increase the PoF value.

In Appendix B of Enclosure 1 to the September 9, 2020, supplement (the information repeated in Tables 2 and 3 of SIA Report No.1900064.406.R0), the licensee provided the examination coverage of the Vogtle requested SG nozzles. The licensee showed that the FW NIR (Examination Category C-B, Item No. C2.22) received 100-percent coverage during the all three ISI examinations. Thus, for the FW nozzle NIR, the assumption of 100-percent examination coverage during all specified times of ISI in the PFM evaluations is reasonable.

For the SG MS and FW nozzle NSWs (Examination Category C-B, Item No. C2.21), the licensee showed that the examination coverage could be as low as 50 percent. In Enclosure 2 to the submittal (SIA Report No.1900064.406.R0), the licensee showed the effect of 50-percent examination coverage for the limiting case in Table 8-9 of the EPRI report, FEW-P3A. The licensee assumed the POD curve (based on ultrasonic testing) in the EPRI report for the PSI and ISI.

The licensee compared this POD curve with a curve based on radiographic testing that is typically used during PSI, and the comparison showed that the POD curve in the EPRI report was more conservative. The licensee assumed 100-percent coverage during PSI based on the collective initial ASME Code,Section III fabrication and initial ASME Code,Section XI examinations. The NRC staff finds for the Vogtle plant-specific relief request that the assumptions described above are acceptable.

The licensee then applied the ASME Code,Section XI, required ISI schedule and compared the PoF results with those from applying the proposed alternative ISI schedule of PSI + 10 + 20 +

30 + 60. The comparison (Table 4 of SIA Report No. 1900064.406.R0) showed that the probability of rupture at 80 years between 100-percent examination coverage and 50-percent examination coverage stayed the same at 1.25x10-9 per year for both ISI scenarios. The comparison also showed that for the ASME Code,Section XI, ISI scenario, the probability of leakage at 80 years increased from 1.25x10-9 per year to 5.93x10-6 per year.

For the licensees proposed alternative of PSI + 10 + 20 + 30 + 60, the probability of leakage at 80 years increased from 2.50x10-9 per year to 5.95x10-6 per year. The NRC staff noted the significant increase in probability of leakage going from 100-percent to 50-percent examination coverage. The probability of leakage increased to a value greater than 1x10-6 per year, but the NRC staff determined that leakage is not component rupture and would be managed by the plant leakage detection system.

Based on the above, the NRC staff finds for thie Vogtle plant-specific relief request in the licensees supplement dated September 9, 2020, adequately addresses the effect of ISI schedule and examination coverage on the PoF values for the requested SG nozzles.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8 Other Considerations 3.8.8.1 Initial Flaw Depth and Length Distribution In Section 8.2.2 of the EPRI report, EPRI stated that the statistical distribution of initial flaw depth used in the PFM analyses is based on the data from the PWR vessel used in the PVRUF project. The NRC staff noted that the PVRUF depth distribution consisted of mostly small fabrication flaws that break the inner surface of the component. Thus, the initial flaws in the PFM analyses consist of small surface breaking flaws.

The flaw data on which the PVRUF depth distribution are based have been used extensively in PFM analyses of PWR components that have been accepted by the NRC, notably in the development of the technical basis 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 [ADAMS Accession No. ML072830074]). Considering this and the discussion in Section 3.8.5 of this SE of flaw density and thickness as they relate to Vogtle, the NRC staff determined for the Vogtle plant-specific relief request that applying the PVRUF depth distribution to the SG vessel locations - the SG MS NSW and SG FW NSW and NIR, is reasonable.

In Section 8.2.2 of the EPRI report, EPRI also described the length distribution used in the PFM analyses. EPRI cited NUREG/CR-6817, A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code, dated April 2004 (ADAMS Accession No. ML040830499), and a proprietary SIA document, 1700313.301, which specifically describes the development of the log-normal distribution for the length. As the NRC staff observed in Item 2.b.iv of the audit report, the flaw data for the length distribution was derived from the most conservative of three sets of flaw data, and as such, the NRC staff finds for the Vogtle plant-specific relief request that the length distribution is acceptable for the analysis results applicable to the requested SG nozzles of Vogtle.

3.8.8.2 Probability of Detection In Section 8.2.2.3 of the EPRI report, EPRI stated that the POD curve used in the analyses was the same POD curve used in the BWRVIP-108 analyses. The NRC staff confirmed that the POD curve in Figure 8-2 in the EPRI report is the same as the POD curve in BWRVIP-108. The NRC staff noted that the NSWs and NIR sections analyzed in BWRVIP-108 were associated with the RPV and that the POD curve was, therefore, developed based on the ultrasonic testing (UT) requirements in the ASME Code,Section XI, Appendix VIII (this is also reflected in the discussion of POD in the December 19, 2007, SE of BWRVIP-108).

The NSWs and NIR sections in the EPRI report are associated with the SG vessel for which the UT requirements of ASME Code,Section V, apply. The NRC staff noted that, in practice, the POD curve based on the UT requirements of the ASME Code,Section V, could be lower than the POD curve based on the UT requirements of the ASME Code,Section XI, Appendix VIII.

Since the ASME Code,Section V, is used for the UT examination of SG vessel components, applying the ASME Code,Section XI, Appendix VIII-based POD curve to the PFM analyses of SG vessel components could be nonconservative.

In its supplement dated November 23, 2020, the licensee stated that for the requested SG MS and FW NVWs, the examinations are performed in accordance with the ASME Code,Section V, Article 4, and thus, the Appendix VIII-based POD curve may not be appropriate for these welds.

To address this, the licensee referred to the conclusion in the EPRI report that the requested SG MS and FW NVWs and FW NIR are acceptable for the proposed alternative of PSI + 10 +

20 + 30 + 60 based on the PSI-only examination case PFM analyses, stating that any examinations after PSI would only serve to support the PSI-only based PFM results. In the first paragraph of Section 3.8.9 of this SE (and in a brief reference in Section 3.8.7 of the SE), the NRC staff explained its non-acceptance of the licensees conclusion on PSI-only examination.

To evaluate the acceptability of the Appendix VIII-based POD curve on the requested SG MS and FW NVWs, the NRC staff assessed the PVRUF cumulative probability distribution shown by the Equation 8-1 of the EPRI report against the Appendix VIII-based POD curve shown in Figure 8-2 of the EPRI report. The PVRUF distribution represented by Equation 8-1 of the EPRI report, in effect, says that there is about a 90-percent probability that the initial flaw depth used in the PFM analyses is equal to or less than 0.0787 inches.

This flaw depth is on the lower portion (left side) of the Appendix VIII-based POD curve in Figure 8-2. While the NRC staff expects that, in practice, a POD curve based on the ASME Code,Section V, could be lower than the Appendix VIII-based POD curve, it would not be much lower for flaw depths equal to or less than 0.0787 inches and flaw depths that are analyzed 90 percent of the time and for which the POD is already very low at about 18 percent.

Based on the above and that POD is not one of the parameters that significantly affects the PFM results (see Section 3.8.2 of this SE), the NRC staff determined for the Vogtle plant-specific relief request that a Section V-based POD curve would have minimal impact on the PFM results compared to an Appendix VIII-based POD curve. Therefore, the NRC staff finds for the Vogtle plant-specific relief request that the Appendix VIII-based POD curve is adequate for use in the PFM analyses for the requested SG MS and FW NVWs.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8.3 Models The NRC staff evaluated the flaw distribution and POD models in Sections 3.8.8.1 and 3.8.8.2, respectively, of this SE and the fracture toughness and the FCG rate models in Sections 3.8.4 and 3.8.6, respectively.

In Section 8.2.2.5 of the EPRI report, EPRI described the fracture mechanics models used in these analyses. For both semi-elliptical circumferential and axial surface cracks in pipe, EPRI employed SIF models from API-579/ASME-FFS-1. EPRI used these SIF models to analyze flaws in the NSW.

The NRC staff questioned the use of cylindrical flaw models for flaws in nozzle welds, and as indicated in response to Item 1.b of the audit report, the licensee stated that the cylindrical SIFs were demonstrated to be conservative with respect to the geometric configuration of the NSW.

Similar models were used to analyze flaws in the NSWs in BWRVIP-108 and BWRVIP-241, which the NRC staff approved in SEs dated December 19, 2007, and April 19, 2013 (ADAMS Accession No. ML13071A240), respectively. Since the Vogtle SG NSWs are similar in configuration to the components analyzed in BWRVIP-108 and BWRVIP-241, the NRC staff finds for this Vogtle plant-specific relief request that the cylindrical SIF models used for the NSWs are acceptable for the analysis results applicable to the subject SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use.

For the NIR crack model, EPRI employed a weight function-based SIF solution. Since this is a unique solution, the NRC staff questioned the V&V of the approach. The NRC staff observed during the audit, as described in Item 1.b of the audit report, the licensees V&V effort for the NIR crack model. The NRC staff finds the use of the weight function-based SIF solution for the NIR crack cases acceptable for the Vogtle plant-specific relief request because it provided similar SIF results as compared to finite element results. The NRC has neither reviewed nor approved the EPRI report for generic use.

The NRC staff further observed (Item 1.b of the audit report) that the NIR crack model is set to a single depth-to-length ratio of 0.5, and thus, there is no length distribution for the NIR crack model unlike the crack model used for the NSW. The NRC staff finds for the Vogtle plant-specific relief request that the single depth-to-length ratio of 0.5 for the NIR crack is acceptable because this ratio has been accepted for postulated flaws, namely, the flaw model used for the nozzle corner flaw in Appendix G to Section XI of the ASME Code.

Based on the above, the NRC staff finds for the Vogtle plant-specific relief request that the SIF models for the NIR sections are acceptable for the analysis results applicable to the requested SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8.4 Uncertainty In Section 8.2.1.2 of the EPRI report, EPRI considered both aleatory uncertainty (random or inherent uncertainty) and epistemic uncertainty (uncertainty due to state of knowledge) and stated that this entailed two sampling loops: an aleatory and epistemic loop. In Section 8.2.4.1, EPRI stated that it considered all random parameters aleatory because they are conservative or based on large sets of data (for example, the fatigue crack growth distribution was developed from over 1,000 fatigue datapoints developed in PWR water environments), and unless otherwise stated, performed 10 million aleatory realizations and 1 epistemic realization.

The NRC staff notes that representing all variables as aleatory will result in probabilities that represent the mean of the distribution.

The NRC staff documented observations on percent error and implementation of aleatory and epistemic realizations in the audit report (Items 1.c and 2.a.ii). With regard to the observation on percent error, the NRC staff notes that large percent errors that result from probabilistic analyses where only one failure happens in 10 million realizations can be impactful if the results approach the acceptance criteria. Assuring sufficient realizations and proper sampling of the input space will reduce the error with these calculations. In addition, the overuse of conservative inputs in a probabilistic analysis can mask the importance of other random valuables and should be avoided. However, since the limiting location for the base case had probabilities of leakage and rupture more than two orders of magnitude below the acceptance criteria, the NRC staff finds for the Vogtle plant-specific relief request that the large uncertainty in the low probability results reflected by the large percent error reasonable.

With regard to the observation on implementation of aleatory and epistemic realizations, the NRC staff notes that from the distribution of results presented, the 50th percentile (median) was very close to the results when the licensee assumed all aleatory realizations (mean). This is expected for a distribution that is normally distributed but may be different for a skewed distribution.

In Section 8.2.2.4.1 of the EPRI report, EPRI stated that transient stresses are normally distributed. The NRC staff noted that this is consistent with Table 8-3 of the EPRI report, which indicates a normal distribution for transient stresses. However, in the list of inputs for the PFM base cases in Table 8-7 of the EPRI report, EPRI stated that the uncertainties on transients are None, which implies that there is no statistical distribution on transient stresses.

During the audit of PROMISE Version 1.0, the NRC staff reviewed one of the input files, which seemed to indicate that the transient stresses have a statistical distribution (Item 2.h.i of the audit report). However, during the audit discussion, the licensee stated that the stresses (which the NRC staff implies include transient stresses) were constant. In its supplement dated November 23, 2020, the licensee clarified that transient stresses were treated as a constant input in the PFM assessments in the EPRI report. The NRC staff finds for the Vogtle plant-specific relief request that treating transient stresses as constant rather than random is acceptable since the transients were selected based on large temperature and pressure variations, as discussed in Section 3.8.3.2 of this SE.

Uncertainties on other random parameters not covered in this section are discussed in other sections of this SE (e.g., Section 3.8.6 for FCG rate).

Based on the above and the observations in the audit report, the NRC staff finds for the Vogtle plant-specific relief request that the licensees handling of uncertainty is acceptable for the analysis results applicable to the requested SG nozzles of Vogtle.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8.5 Software V&V In Section 8.2.3.2 of the EPRI report, EPRI discussed the two phases that comprised the V&V of PROMISE Version 1.0: a testing phase and a benchmarking phase. The testing phase was based on a software V&V plan (SVVP) and the results of the testing documented in a V&V report; both the SVVP and the V&V report are proprietary SIA reports. The benchmarking phase included comparison of probability of leak values with VIPERNOZ, another SIA PFM software.

The NRC staff reviewed the benchmarking activities described in Section 8.2.3.2.2 of the EPRI report and determined that, together with the benchmarking of ISI in Enclosure 3 to the September 9, 2020, supplement (1900064.407.R2) discussed in Section 3.8.7 of the SE, the benchmarks EPRI performed for PROMISE Version 1.0 were adequate for the Vogtle plant-specific relief request.

Based on the above and the NRC staffs observations in the audit report on V&V (Items 1.a and 1.b), the NRC staff finds for the Vogtle plant-specific relief request that PROMISE Version 1.0 received adequate V&V, and therefore, is acceptable for use in the licensees alternative request for the requested SG nozzles.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8.6 Convergence In Section 8.2.4.3.10 of the EPRI report, EPRI conducted a SS to determine if the number of realizations chosen for this effort resulted in a converged solution. For this study, EPRI chose case FEW-P3A (limiting FW nozzle location) with a stress multiplier of 1.5 and conducted analyses with both 107 and 108 realizations. EPRI illustrated in Table 8-27 of the EPRI report that the leakage probabilities calculated were insensitive to the increase in realizations.

Based on the above and the NRC staffs observation in the audit report on convergence (Item 2.f.i), the NRC staff finds for the Vogtle plant-specific relief request that the number of realizations used in the analyses is acceptable for the results applicable to the requested SG nozzles of Vogtle, even though, as described in Section 3.8.8.4 of this SE, the uncertainty is high for those cases where only one failure occurs within an analysis.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.8.7 DFM Analysis In Section 8.3 of the EPRI report, EPRI performed a DFM analysis with an initial flaw depth of approximately 5 percent of the component thickness and average values of all other parameters considered random in the PFM analysis. The initial flaw depth is obtained from the ASME Code,Section XI, Table IWC-3510-1.

All analyzed locations in the subject SG nozzles resulted in many years to reach leakage (80 percent of the component thickness), the least being 147 years. No locations reached an applied SIF of greater than 200 ksiin, but the FW nozzle NIR resulted in an applied SIF of 101 ksiin at 80 years, which, considering a factor safety of 2, is greater than the allowable SIF value of 200/2 = 100 ksiin. The NRC staff finds this acceptable for Vogtle since NIR sections received 100% coverage with no recordable indications, and the results of the DFM are for a postulated flaw. The NRC staff also noted that the results for the FW nozzle NIR in the EPRI report are slightly conservative because the applied loads include the effect of welding residual stress (see Section 3.8.3.3 of this SE). Thus, the NRC staff determined for the Vogtle plant-specific relief request that overall, the DFM analysis supports the PFM analysis.

The NRC has neither reviewed nor approved the EPRI report for generic use.

3.8.9 PFM Results Relevant to Alternative VEGP-ISI-ALT-04-04 In Section 6.0 of Enclosure 1 to its supplement dated the September 9, 2020, the licensee stated that based on the PFM results, after PSI, no other inspections are required for up to 60 years of plant operation to meet the NRCs safety goal of 1x10-6 failures per year. A similar observation is noted in Section 8.2.4.1.1 of the EPRI report, which states that performing only PSI examination without any other post-PSI examinations is acceptable for 80 years of plant operation while maintaining plant safety.

The NRC staff does not find either general conclusion by the licensee and the EPRI report acceptable since it does not account for the effect of the combination of the most significant parameters of stress, fracture toughness, and flaw density or the added uncertainty of low probability events. The NRC staff determined that since the PFM analyses were based on representative SG MS and FW nozzles, the uncertainties on the different parameters (which are different from the sampling uncertainty discussed in Section 3.8.8.4 of this SE) should be taken into account, especially the significant parameters of stress, fracture toughness, and flaw density, before a general conclusion can be made. As an example, even for a case more favorable than PSI only, such as PSI followed by 20-year ISI (PSI + 20 + 40 + 60 or PSI, 20 year interval in Figure 8-9 or 0, 20, 40, 60 in Table 8-10 of the EPRI report), the probability of rupture at 80 years for the FW nozzle NIR changes from 1.25x10-12 per year to 5.3x10-6 per year (see Table 8-28 of the EPRI report), which is above the criterion of 1x10-6 per year. This was a result of a SS that changed at the same time stress, fracture toughness, and flaw density (at the NIR) to more conservative values, which reasonably accounted for uncertainties.

While the NRC staff acknowledged that this study assumed conservative values for all three parameters simultaneously, the NRC staff also noted that had the same study been performed for the PSI-only case, the probability of rupture values would have been much higher, and that only one or two parameters with conservative values could easily lead to probability of ruptures greater than 1x10-6 per year.

In addition, as described in Section 3.8.8.4 of this SE, the NRC staff calculated a large percent error when using the procedure described in the EPRI report. As described in the NRC audit report, the licensee explained that the larger error is expected and should have minimal impact since the resulting probabilities are so low. However, results for the cases where only PSI was considered are much closer to the acceptance criteria; therefore, these errors need to be addressed through sufficient realizations and proper sampling of the input space before general conclusions may be considered for the PSI-only cases.

Lastly, PSI-only examinations, as compared to the proposed alternative of PSI + 10 + 20 + 30 +

60, would have a much more adverse effect on risk-informed principles, particularly since PSI-only examinations would remove any future condition monitoring needed for risk-informed decision making.

As previously discussed in Section 3.8.7 of this SE, the licensee is seeking the alternative ISI schedule of PSI + 10 + 20 + 30 + 60 (equivalent to PSI + 10 + 20 + 30, 30 year interval in Figure 8-9 or 0, 10, 20, 30, 60 in Table 8-10 of the EPRI report). Therefore, the NRC staff determined that PFM results for PSI + 10 + 20 + 30 + 60 are the results relevant to the licensees proposed alternative, as the NRC staff discusses next.

The NRC staff noted that the EPRI report does not have PoF results for PSI + 10 + 20 + 30 +

60, but also noted that the ISI scenario PSI + 20 + 40 + 60 has similar PoF results, as shown in Figure 8-9 and Table 8-10 of the EPRI report. The PoF results in Table 8-8 of the EPRI report for PSI + 20 + 40 + 60 show a large margin from the criterion of 1x10-6 per year. However, Table 8-8 applies only for base cases that have base-case inputs, and, as the NRC staff discussed in the first paragraph of this section, sensitivity studies reasonably address the uncertainties due to the various input parameters. The NRC staff, therefore, evaluated the results in the sensitivity studies in Section 8.2.4.3 of the EPRI report relevant to the proposed alternative ISI schedule of PSI + 10 + 20 + 30 + 60.

As noted previously in this section, the probability of rupture for the combination case in Table 8-28 of the EPRI report for PSI + 20 + 40 + 60 (again, similar to PSI + 10 + 20 + 30 + 60) is 5.30x10-6 per year for the FW nozzle NIR, which is the limiting case. The value of 5.30x10-6 per year is greater than 1x10-6 per year, and the parameters changed were the following: stress (multiplier of 1.0 to 1.5), standard deviation on upper shelf KIC (5 ksiin to 30 ksiin), and flaw density (0.001 to 0.1 flaw per nozzle). In Section 3.8.4 of this SE, the NRC staff discussed the mean and standard deviation values of upper shelf KIC used in the EPRI report and found them acceptable. Therefore, the NRC staff evaluated the effect of the other two parameters - stress and flaw density, assuming mean upper shelf KIC and its standard deviation were the same as the base case.

Tables 8-15 and 8-16 of the EPRI report contain the SS on stress. Table 8-15 shows that the probabilities of rupture are well below the criterion of 1x10-6 per year, even when considering the 0.1 flaw per nozzle at the FW nozzle NIR, which would change the probability of rupture results for FEW-P1N and FEW-P2N from 1.25x10-12 to 1.25x10-10 per year (see discussion in Section 3.8.5 of this SE on NIR flaw density as a multiplier).

Table 8-16 shows that limiting probabilities of leakage are 1.04x10-6 per year at the FW nozzle NSW and 1.06x10-6 per year at the FW NIR (adjusted for 0.1 flaw per nozzle). Even though these probabilities are greater than 1x10-6 per year, the NRC staff finds them acceptable for the Vogtle plant-specific relief request because they are leakage probabilities as opposed to rupture probabilities.

The NRC staff also noted that the results for the FW nozzle NIR are slightly conservative because the applied loads include the effect of welding residual stress (see Section 3.8.3.3 of this SE). Finally, the NRC staff noted since the licensees proposed alternative is through 60 years of operation (through the sixth ISI interval as stated in Section 7.0 of Enclosure 1 to the supplement dated September 9, 2020), the probability values should be based on 60 years of operation.

Tables 8-15 and 8-16 of the EPRI report are for 80 years of operation, and at 60 years of operation, the results could be up to 80/60 = 1.3 times larger since the number of failures would be divided by 60 years instead of 80 years (assuming the number of failures have been reached by 60 years). The NRC staff determined that this factor of 1.3 has no impact on the NRC staffs conclusion on the probabilities shown in Tables 8-15 and 8-16 of the EPRI report. Thus, the NRC staff determined for the Vogtle plant-specific relief request that the PFM analyses in the EPRI report adequately address uncertainties in the PoF values relevant to the licensees proposed alternative of PSI + 10 + 20 + 30 + 60.

Based on the above, the NRC staff finds for the Vogtle plant-specific relief request that the proposed alternative of PSI + 10 + 20 + 30 + 60 for the requested SG nozzles of Vogtle would result in a PoF per year that is reasonably below the criterion of 1x10-6 per year.

The NRC has neither reviewed nor approved the EPRI report for generic use.

4.0 CONCLUSION

As set forth above, the NRC staff determined that the licensees proposed alternative in VEGP-ISI-ALT-04-04 to increase the ISI interval from 10 years to 30 years 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 in VEGP-ISI-ALT-04-04 for Vogtle, Units 1 and 2, for the duration of the fourth 10-year ISI interval through the sixth 10-year ISI interval.

The NRC did not review the EPRI report for generic use, and this approval does not extend beyond the Vogtle plant-specific authorization.

All other requirements of Section XI of the ASME Code for which an alternative was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: D. Dijamco D. Rudland Date: January 11, 2021

ML20352A155 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME JLamb KGoldstein (LRonewicz for)

HGonzalez DATE 12/18/2020 12/18/2020 12/15/2020 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/BC NAME JLamb MMarkley DATE 12/19/2020 01/11/2021