ML23073A191

Alternative for Inservice Inspections of Pressurizer and Steam Generator Weld Examinations

ML23073A191
Person / Time
Site:	Surry
Issue date:	04/21/2023
From:	Markley M NRC/NRR/DORL/LPL2-1
To:	Stoddard D Dominion Nuclear
References
EPID L-2022-LLR-0081
Download: ML23073A191 (1)

Text

April 21, 2023 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2 - RE: ALTERNATIVE REQUEST TO DEFER ASME CODE SECTION XI EXAMINATIONS OF PRESSURIZER AND STEAM GENERATOR WELDS AND NOZZLES (EPID L-2022-LLR-0081)

Dear Mr. Stoddard:

By letter dated November 17, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22322A158), as supplemented by letters dated February 11 and March 2, 2023 (ML23044A039, and ML23061A108, respectively), Virginia Electric and Power Company (the licensee, Dominion), Dominion Energy Virginia submitted a request for Surry Power Station (Surry) Units 1 and 2, to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI examinations.

Specifically, Dominion requested to defer selected pressurizer and steam generator weld examinations through the sixth 10-year ISI interval ending on October 13, 2033, for Surry Unit 1, and through the remainder of the third period of the fifth 10-year ISI interval through the sixth 10-year ISI interval ending on May 9, 2034, for Surry Unit 2, on the basis that the proposed alternative provides an acceptable level of quality and safety. The NRC staff reviewed the proposed alternative request for Surry Units 1 and 2, as a plant-specific alternative. The NRC did not review the EPRI reports for generic use, and this approval does not extend beyond the Surry, Units 1 and 2, plant-specific authorization.

The NRC staff has reviewed the alternative request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CRR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative through the sixth 10-year ISI interval ending on October 13, 2033, for Surry Unit 1, and through the remainder of the third period of the fifth 10-year ISI interval through the sixth 10-year ISI interval ending on May 9, 2034, for Surry Unit 2. All other requirements of ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

D. Stoddard If you have any questions, please contact the Project Manager, John Klos at 301-415-5136 or via email at John.Klos@nrc.gov.

Sincerely, Michael T. Digitally signed by Michael T. Markley Date: 2023.04.21 09:50:47 -04'00' Markley Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.

50-280 50-281

Enclosure:

Safety Evaluation cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST TO DEFER ASME CODE SECTION XI EXAMINATION OF PRESSURIZER AND STEAM GENERATOR WELDS AND NOZZLES WELD INSPECTIONS VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNITS 1 AND 2 DOCKET NO.S 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated November 17, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22322A158), as supplemented by letters dated February 11 and March 2, 2023 (ML23044A039, and ML23061A108, respectively), Virginia Electric and Power Company (the licensee, Dominion), Dominion Energy Virginia submitted a request for the Surry Power Station (Surry), Units 1 and 2, to the U.S. Nuclear Regulatory Commission (NRC or Commission) for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, inservice inspections (ISI).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to defer selected Pressurizer (PZR) and Steam Generator (SG) weld examinations through the sixth 10-year ISI interval ending on October 13, 2033, for Surry Unit 1, and through the remainder of the third period of the fifth 10-year ISI interval through the sixth 10-year ISI interval ending on May 9, 2034, for Surry Unit 2, on the basis that the proposed alternative provides an acceptable level of quality and safety. In its submittal, the licensee requested the proposed alternative on the basis of plant-specific applicability of three technical reports prepared by the Electric Power Research Institute (EPRI). The NRC staff reviewed the proposed alternative request for Surry as a plant-specific alternative. The NRC did not review the EPRI reports for generic use, and this approval does not extend beyond the Surry, Units 1 and 2, plant-specific authorization.

2.0 REGULATORY EVALUATION

The PZR and SG pressure-retaining welds and SG full penetration welded nozzles at Surry, Units 1 and 2, are ASME Code Class 1 and 2 components, whose ISI are performed in accordance with Section XI, Rules for Inservice Inspection of Nuclear Power Plant Enclosure

Components, of the ASME Code and applicable edition and addenda, as required by 10 CFR 50.55a(g).

The regulations in 10 CFR 50.55a(g)(4) state, in part, components that are classified as ASME Code Class 1, 2, and 3 must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a may be used when authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the alternative and the NRC staff to authorize it.

3.0 TECHNICAL EVALUATION

3.1 Licensees Proposed Alternative

Applicable Code Edition and Addenda

The ASME Code of record for the Surry, Units 1 and 2, fifth 10-year inservice inspection (ISI) interval is the ASME Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, 2004 Edition. The fifth 10-year ISI interval for Surry, Unit 1, started on October 14, 2013, and ends on October 13, 2023. The fifth 10-year ISI interval for Surry, Unit 2, started on May 10, 2014, and ends on May 9, 2024. The incorporation by reference of the 2019 Edition of ASME Code,Section XI in 10 CFR 50 was published as a FRN in the Federal Register on October 27, 2022 at 87 FR 65128, with an effective date of November 28, 2022. Thus, in accordance with 10 CFR 50.55a(g)(4)(iii), which states that 10-year ISI intervals subsequent to the first 10-year ISI interval must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(a) 18 months before the start of the 10-year ISI interval. Thus, the ASME Code of record for the sixth 10-year ISI interval for Surry, Unit 1 (October 14, 2023, to October 13, 2033) and Unit 2 (May 10, 2024, to May 9, 2034) would be the 2019 Edition of ASME Code,Section XI.

ASME Code Components Affected ASME Code Class: Section XI, Class 1 and 2 Examination Category: B-B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels C-A, Pressure-Retaining Welds in Pressure Vessels C-B, Pressure-Retaining Nozzle Welds in Pressure Vessels Item Numbers: B2.11 for PZR circumferential welds B2.12 for PZR longitudinal welds B2.40 for SG vessel primary side, tubesheet-to-head welds

C1.10, C1.20, and C1.30 for SG vessel secondary side welds C2.21 for the nozzle-to-shell weld C2.22 for the nozzle inside radius (NIR) section Component IDs: Table 2 of Attachment 1 to the initial submittal (shown below) lists the component identifications (IDs) affected.

ASME Code Requirements for Which Alternative is Requested For ASME Code Class 1 welds in the PZR, the ISI requirements are those specified in Subarticle IWB-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric examinations as specified in ASME Code,Section XI, Table IWB-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval:

Examination Category B-B, Item No. B2.11, PZR Shell-to-Head Welds, Circumferential Examination Category B-B, Item No. B2.12, PZR Shell-to-Head Welds, Longitudinal For ASME Code Class 1 welds in the steam generator (SG), the ISI requirements are those specified in Subarticle IWB-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric examinations as specified in ASME Code,Section XI, Table IWB-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval. As noted in Table IWB-2500-1 for Examination Category B-B, cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.

Examination Category B-B, Item No. B2.40, SG Primary Side Tubesheet-to-Head Welds For ASME Code Class 2 welds and nozzle inside radius (NIR) sections in the SG, the ISI requirements are those specified in Subarticle IWC-2500 of the ASME Code,Section XI, which requires the licensee to perform volumetric and surface examinations as specified in ASME Code,Section XI, Table IWC-2500-1, for each Examination Category and Item No. listed below once every 10-year ISI interval. As noted in Table IWC-2500-1 for Examination Categories C-A and C-B, cases of multiple vessels of similar design, size, and service (such as SGs), the required examinations may be limited to one vessel or distributed among the vessels.

Examination Category C-A, Item No. C1.10, Shell Circumferential Welds Examination Category C-A, Item No. C1.20, Head Circumferential Welds Examination Category C-A, Item No. C1.30, Tubesheet-to-Shell Welds Examination Category C-B, Item No. C2.21, Nozzle-to-Shell Welds Examination Category C-B, Item No. C2.22, Nozzle Inside Radius Sections

As discussed previously in this safety evaluation (SE), the 2019 Edition of ASME Code,Section XI has been incorporated by reference in 10 CFR 50.55a(a)(1)(ii). The NRC confirmed that the ASME Code requirements listed above did not change going from the 2004 Edition to the 2019 Edition of the ASME Code,Section XI.

Reason for Proposed Alternative In its letter dated November 17, 2022, Attachment 1, Section 4.0, Reason for Request, the licensee stated that the Electric Power Research Institute (EPRI) performed non-proprietary reports/assessments which formed the basis for the ASME Code,Section XI PZR and SG welds/component examination requirements. Those reports are listed below and apply to the components listed in Section 3.1 above.

EPRI Technical Report 3002015905, Technical Bases for Inspection Requirements for PWR [Pressurized Water Reactor] Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, 2019 (here after referred to as EPRI report 15905, ML21021A271).

EPRI Technical Report 3002015906, Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 2019 (here after referred to as EPRI report 15906, ML20225A141).

EPRI Technical Report 3002014590, Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, 2019 (here after referred to as EPRI report 14590, ML19347B107).

The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The licensee stated that these reports were developed consistent with EPRIs White Paper on PFM, White Paper on Suggested Content for PFM Submittals to the NRC, (ML19241A545) and that the reports concluded that the current ASME Code,Section XI ISI interval of 10 years can be increased significantly with no impact to plant safety. Based on the conclusions of the three reports, the licensee is requesting a plant-specific alternative to the 10-year ISI interval for the subject welds.

The NRC staff noted that the EPRI reports were not submitted or reviewed as a topical report.

The NRC staff reviewed the proposed alternative request for Surry, Units 1 and 2, as a plant-specific alternative. The NRC did not review the EPRI reports for generic use, and this review does not extend beyond the Surry, Units 1 and 2, plant-specific authorization.

Proposed Alternative and Basis for Use In Section 5.0 of Attachment 1 to its submittal, the licensee stated that the proposed alternative increases the ISI interval for these examination items from the current ASME Code,Section XI, 10-year requirement thereby deferring examinations for the remainder of the current sixth 10-year ISI interval for Unit 1, for the remainder of the fifth and sixth 10-year ISI interval for Unit 2. The licensee stated that some components have not yet received an ISI examination for the fifth 10-year ISI interval. By letter dated February 11, 2023, the licensee supplemented its submittal in responding to the NRC request for additional information. Section 3.2.11 of this SE discusses the licensees performance monitoring.

Duration of Proposed Alternative The licensee requested that the proposed alternative increases the inspection interval for examination items addressed in the submittal from the current ASME Code Section XI 10-year requirement, thereby deferring the PZR and SG examinations for the following:

Surry Unit 1 through the sixth 10-year ISI interval, which ends on October 13, 2033, and Surry Unit 2 for the remainder of the third period of the fifth 10-year ISI interval through the sixth 10-year ISI interval, which ends on May 9, 2034.

Basis for Proposed Alternative In its letter dated November 17, 2022, Attachment 1, Section 5.0, Proposed Alternative and Basis for Use, the licensee discussed the key aspects of the technical basis in the EPRI report and its applicability tom Surry, Units 1 and 2. EPRI report 15905 was used as basis for proposed alternative for the PZR ASME Code Examination Categories B-B welds. EPRI report 15906 was used as basis for proposed alternative for the SG ASME Code Examination Categories B-B and C-A welds. EPRI report 14590 was used as basis for proposed alternative for the SG ASME Code Examination Category C-B welds and NIR.

The NRC staffs review focused on evaluating the applicability of the PFM analyses in Section 8.3 of EPRI report 15905 and EPRI report 15906, and Section 8.2 of EPRI report 14590, and verifying whether the DFM and PFM analyses in the reports support the proposed alternative.

The licensee cited NRC-approved precedents for its request that were based on EPRI reports 15905, 15906, and 14590. These precedents included Salem Generating Station, Units 1 and 2, submittal (ML20218A587, hereafter Salem submittal), a Millstone Power Station Unit 2 submittal (ML20198M682, hereafter Millstone submittal), and a Vogtle Electric Generating Plant, Units 1 and 2, submittal (ML19347B105, hereafter Vogtle submittal). The licensee referenced applicable portions of the technical arguments from these submittals. The NRC staff documented its review of these applications in the associated plant-specific SEs (Salem (ML21145A189), Millstone (ML21167A355), and Vogtle (ML20352A155). For the Surry, Units 1 and 2, review, the NRC staff considered the information referenced and focused on the plant-specific application of the EPRI reports for Surry, Units 1 and 2. Consistent with the key principles of the NRC risk-informed approach, the NRC staff also confirmed that the proposed alternative provides sufficient performance monitoring.

3.2 NRC Staff Evaluation 3.2.1 Degradation Mechanisms In Section 5.0 of Attachment 1 to its submittal, the licensee stated, in part, that:

An evaluation of degradation mechanisms that could potentially impact the reliability of the PZR and SG welds and components was performed in References [1-2], [1-3], and

[1-4]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue.

Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the PZR and SG welds and components covered in this request. Therefore,

only these fatigue-related mechanisms considered in the PFM and DFM evaluations in References [1-2], [1-3], and [1-4] are applicable to the components in this request.

The NRC staff reviewed the submittal for plant-specific circumstances that may indicate presence of a degradation mechanism and activity sufficiently unique to Surry, Units 1 and 2, to merit additional consideration. Such circumstances pertain to materials of the subject components, stress states, and reactor coolant environment. The NRC staff found no evidence of conditions at Surry, Units 1 and 2, that would require consideration of a unique degradation mechanism beyond application of EPRI reports 15905, 15906, and 14590. Specifically, the NRC staff reviewed the materials, stress states, and chemical environment (i.e., reactor coolant) of the subject PZR welds, and SG welds and NIR, and found them to be consistent with the assumptions made in the EPRI reports. Therefore, the NRC staff finds that consideration of additional degradation mechanisms beyond those from the EPRI reports is not necessary.

3.2.2 PFM Analysis In Section 5.0 of Attachment 1 to its submittal, the licensee stated, in part, that:

Finite element analyses (FEA) were performed in References [1-2], [1-3], and [1-4] to determine the stresses in the PZR and SG welds and components covered in this request. The finite element models used in References [1-2], [1-3], and [1-4] are consistent with the configurations of SPS, Units 1 and 2, therefore no new FEA model is required for the stress analysis of these units. The analyses were performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to SPS, Units 1 and 2, is demonstrated in Attachments 2 and 3 and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions contained in References

[1-2], [1-3], and [1-4] are applicable to SPS, Unit 1 and Unit 2. In particular, the key geometric parameters used in the stress analyses in References [1-2], [1-3], and [1-4]

are compared to those of SPS, Units 1 and 2, in Table 3 for the PZR and Tables 4 and 5 for the SGs:

The licensee also stated, in part, that:

Flaw tolerance evaluations were performed in References [1-2], [1-3], and [1-4]

consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent ISI, the NRCs safety goal of 1x10-6 failures per year is met.

The NRC staff confirmed that the analysis provided by the licensee for the Surry, Units 1 and 2, submittal is consistent with the approach taken in the Salem, Millstone, and Vogtle precedents and explicitly referenced for plant-specific applicability in the Surry, Units 1 and 2, request. The NRC staff determined that the PFM analysis is consistent and, therefore, the NRC staff finds the proposed PFM analysis to be appropriate for this application for Surry, Units 1 and 2. The NRC staff evaluated the licensees claim about the impact of PSI on the PFM analysis in Section 3.2.10 of this SE.

The NRC staff noted that the acceptance criterion of 1x10-6 failures per year (also termed Probability of Failure, PoF) is tied to that used by the NRC staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal

shock events, and other similar reviews. In that rule, the reactor vessel through-wall crack frequency (TWCF) of 1x10-6 per year for a pressurized thermal shock event is an acceptable criterion because reactor vessel TWCF is conservatively assumed to be equivalent to an increase in core damage frequency, and as such would meet the guidance in Regulatory Guide (RG) 1.174, An Approach to for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, (ML17317A256). The discussion of TWCF is explained in detail in the technical basis document for 10 CFR 50.61a, NUREG-1806 Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), August 2007 (ML072830074).

The NRC staff also noted that the TWCF criterion of 1x10-6 per year was generated using a conservative model for reactor vessel cracking. In addition, this criterion exists within a context of reactor pressure vessel surveillance programs and inspection programs. The NRC staff finds that the licensees use of 1x10-6 failures per year that is based on the reactor vessel TWCF criterion is reasonable for the requested PZR welds, and SG welds and NIR of Surry, Units 1 and 2, because (a) the impact of a PZR or SG vessel failure would be less than the impact of a reactor vessel failure on overall risk (i.e., meaning contribution to core damage frequency due to a PZR or SG vessel failure would be less than the contribution due to a reactor vessel failure);

(b) the subject welds have substantive, relevant, and continuing inspection histories and programs; and (c) the estimated risks associated with the individual welds are mostly much lower than the system risk criterion (i.e., the system risk is dominated by a small sub-population which can be considered the principal system risk for integrity, which means that failure of an individual weld is likely to lead to only a limited contribution to risk). The NRC staff further noted that comparing the probability of leakage to the same criterion of 1x10-6 failures per year is conservative because leakage is not failure. The use of a PoF criteria such as 1x10-6 per year for individual welds may not be appropriate generically, but based on the discussion above, the NRC staff finds the application of this criterion acceptable for this plant-specific review for the PZR welds, and SG welds and NIR for Surry, Units 1 and 2.

Based on the above, the NRC staff finds the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for the Surry, Units 1 and 2, plant-specific alternative request.

3.2.3 Parameters Most Significant to PFM Results In the following sections, the NRC staff reviewed the following parameters or aspects most significant to the PFM analysis: stress analysis, fracture toughness, flaw density, flaw crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage.

3.2.4 Stress Analysis 3.2.4.1 Selection of Components and Materials In Attachment 2 to the November 17, 2022, submittal, the licensee evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI reports 15905, 15906, and 14590 to the subject PZR welds, and SG welds and NIR of Surry, Units 1 and 2.

These reports evaluated representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The reports also specified plan-specific applicability criteria with regards to component geometries, materials, and loading conditions, that must be evaluated and met by each plant to determine the applicability of the reports. The licensee stated that the plant-specific criteria regarding component geometries and

materials were met. The acceptability of meeting these criteria, however, depends on the acceptability of the component and material selection described in the EPRI reports, which the NRC staff evaluated below. The NRC staff evaluated the loading conditions (i.e., transient selection) criteria in Section 3.2.4.2 of this SE.

In Section 4 of EPRI reports 15905, 15906, and 14590, EPRI discussed the variation among PZR, SG shell, and SG nozzle designs. EPRI used this information for finite element analyses (FEA, see Section 3.2.4.4 of this SE) to determine stresses in the analyzed components, which the licensee referenced for the corresponding PZR and SG components requested for Surry, Units 1 and 2. In selecting the components, EPRI considered geometry, operating characteristics, materials, field experience with respect to service-induced cracking, and the availability and quality of component-specific information.

The NRC staff reviewed Section 4 of EPRI reports 15905, 15906, and 14590, and finds that the PZR and SG configurations selected in the report for stress analysis are acceptable representatives for the corresponding PZR and SG components requested for the Surry, Units 1 and 2, plant-specific alternative request. Specifically, the radius-to-thickness (R/t) ratios of the requested Surry, Units 1 and 2, components, provided in Tables 3 through 5 of Attachment 1 to the submittal, are either bounded by the R/t ratios analyzed in the EPRI reports or such that they are bounded by the SS on stress in the EPRI reports. To verify the dominance of the R/t ratio, the NRC staff reviewed the through-wall stress distributions in Section 7 of the EPRI reports to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. For some of the SG shell welds modeled in EPRI report 15906, the NRC noted that the thermal stress is also potentially high. However, because thickness is the controlling parameter for thermal stress (the lower the thickness, the lower the thermal stress), the NRC staff determined that EPRI report 15906 would still be adequate in providing reasonable assurance for the corresponding welds of Surry, Units 1 and 2, because the thickness values of the Surry, Units 1 and 2, SG vessel, as shown in Table 4 of Attachment 1 to the submittal, are less than those for the SG model analyzed in EPRI report 15906. Accordingly, the NRC staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for the Surry, Units 1 and 2, plant-specific alternative request.

Section 9.4 of EPRI reports 15905, 15906, and 14590 addresses criteria for plant-specific applicability of the analysis and indicates that materials are acceptable if they conform to ASME B&PV Code,Section XI, Paragraph G-2110. The 2019 edition of ASME Code,Section XI, Appendix G, Paragraph G-2110 indicates that Figure G-2210-1 is based on specimens of SA-533 Grade B Class 1, and SA-508-1, SA-508-2, and SA-508-3 steel; thus, the NRC staff noted that additional information for some PZR and SG materials at Surry, Units 1 and 2, were necessary to demonstrate conformance with ASME B&PV Code,Section XI, Nonmandatory Appendix G, paragraph G-2110. The licensee addressed these criteria in Table A1 of to the submittal, and with supplemental information, provided by letter dated February 11, 2023, for Surry, Units 1 and 2.

In particular, the licensee explained, in its supplemental letter dated February 11, 2023, that the applicable construction code for the following components is the 1968 Edition of ASME Section Ill; thus, Subparagraph NB-2331 had not yet been developed:

Pressurizer Shell fabricated from SA-302, Gr. B SG Upper Head and Upper Shell fabricated from SA-533, Gr. A, Cl. 1 SG Feedwater and Main Steam Nozzles fabricated from SA-508, Cl. 2

Pressurizer Upper Head and Lower Head fabricated from SA-216, Gr. WCC Consequently, the licensee provided information for PZR shell, SG upper head and upper shell, and SG feedwater and main steam nozzles in its supplemental letter dated February 11, 2023, to estimate values of initial nil-ductility reference temperature (RTNDT) since measured values are not available due to the applicable construction code (i.e., 1968 Edition of ASME Section Ill).

Additionally, the licensee confirmed that the components were fabricated from materials that have a specified minimum yield strength at room temperature of 50 ksi [kilo-pounds per square inch] less. Based on its review of this information, the NRC staff finds the licensee has demonstrated that criteria 4 of Section B.1.1, Determination of RTNDT for Vessel Materials of BTP 5-3, Revision 3 (ML18254A090), is applicable for these components and that an initial RTNDT of 30°F is reasonable. Additionally, the NRC staff noted that this initial RTNDT of 30°F for the PZR shell, SG upper head and upper shell and SG feedwater and main steam nozzles is bounded by the RTNDT value of 60°F used in the EPRI reports.

With respect to the PZR upper head and lower head, the licensee confirmed that the components were fabricated from materials that have a specified minimum yield strength at room temperature of 50 ksi less. Additionally, the licensee stated that a review of available industry testing data was performed, and determined it was reasonable to assign an RTNDT of 60°F to the SA-216, Gr. WCC, PZR materials fabricated to the 1968 Edition of ASME Section III.

The licensee explained that this determination is based on the relatively high toughness values obtained in the industry testing for similar SA-216, Gr. WCC, materials when compared to KIC values calculated from the ASME Section XI reference curve for an RTNDT of 60°F. The NRC staff noted that the figure provided in licensees supplemental letter dated February 11, 2023, shows that this industry test data is encompassed by the ASME Section XI reference curve for an RTNDT of 60°F. The licensee confirmed that from the available test certificates for the PZR lower head materials reference Westinghouse fabrication and material specification drawings that called for the heat treatments consistent with the industry test data for SA-216, Gr. WCC, discussed above. Additionally, the NRC staff noted that this industry test data was from materials fabricated during the same time period as the Surry PZR materials (i.e., late 1960s or early 1970s) and would have reasonably received heat treatments similar to those specified for the Surry PZR heads. Thus, based on the available industry test data for SA-216, Gr. WCC being encompassed bounded by the ASME Section XI reference curve for an RTNDT of 60°F, the NRC staff finds it reasonable that the initial RTNDT of 60°F for the PZR upper head and lower head is bounded by the RTNDT value of 60°F used in the EPRI reports.

Additionally, the licensee explained in its supplemental letter dated February 11, 2023, that the applicable construction code for the following components is the 1974 Edition through the Winter 1976 Addenda of ASME Section Ill:

SG Lower Head fabricated SA-216, Gr. WCC SG Lower Shell fabricated SA-533, Gr. A, Cl. 2 SG Lower Tubesheet fabricated SA-508, Cl. 2a The licensee confirmed that the construction code for these components required compliance with Paragraph NB-2331, and that the procurement specifications for these components specified a minimum RTNDT of 60°F. Additionally, the licensee confirmed that the SG lower head was fabricated from a material that has a specified minimum yield strength at room temperature of 50 ksi less. Although the materials used to fabricate the SG lower shell and tubesheet (i.e., SA-533, Gr. A, Cl. 2 and SA-508, Cl. 2a, respectively) have a specified minimum yield

strength greater than 50 ksi at room temperature, the NRC staff noted these materials are identified in Table G-2110-1 of the 2019 edition of ASME Code,Section XI, Appendix G; thus, the use of Figure G-2210-1 for these materials is acceptable consistent with the Paragraph G-2110 of the 2019 edition of ASME Code,Section XI, Appendix G [kilo-pounds per square inch]. Thus, based on the construction code for these components and applicable procurement specifications, the NRC staff finds that the initial RTNDT of 60°F for the SG lower head, lower shell and tubesheet is acceptable for use in the Surry plant-specific submittal.

The materials of the Surry, Units 1 and 2, PZR and SGs relevant to the requested welds are specified in Attachment 2 of its submittal and supplemental letter dated February 11, 2023.

Based on its review as described above, the NRC staff verified that these materials at Surry, Units 1 and 2, conform with of Paragraph G-2110 of ASME Code Section XI, Nonmandatory Appendix G; therefore, the NRC staff finds that the materials for Surry, Units 1 and 2, meet the material applicability criterion for the EPRI reports.

Additionally, Table A1, Attachment 2 to the submittal also states that the Surry, Units 1 and 2, PZR, and SG shell and nozzles meet the applicability criteria in EPRI reports 15905, 15906, and 14590 regarding weld and nozzle configuration, attached piping line size, and thermal sleeve attachment. The NRC staff reviewed the licensees information against the applicability criteria and finds that the Surry, Units 1 and 2, PZR and SG meet the applicability criteria described in the EPRI reports.

Based on the above, the NRC staff finds that the licensee has made a plant-specific case that Surry, Units 1 and 2, meets the component geometry and materials applicability criteria in the EPRI reports. The analyzed geometries and materials are acceptable for the requested PZR and SG components at Surry, Units 1 and 2. The NRC concludes that the analyses acceptably model the subject Surry, Units 1 and 2, geometries and materials.

3.2.4.2 Selection of Transients In Attachment 2 to its submittal, the licensee evaluated the plant-specific applicability of the transients selected and analyzed in EPRI Reports 3002015905, 3002015906, and 3002014590 to the subject PZR and SG welds and NIR of Surry, Units 1 and 2. The licensee stated that the plant-specific applicability criteria regarding transients were met. The acceptability of meeting the criteria, however, depends on the acceptability of the transient selection described in the EPRI reports, which the NRC staff evaluated below.

In Section 5.2 of EPRI reports 15905, 15906, and 14590, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to PZRs, SG shell, and SG nozzles. EPRI developed a list of transients for analysis applicable to the PZRs, SG shell, and SG nozzles analyzed in the report, based on transients that have the largest temperature and pressure variations.

The NRC staff evaluated the transient selection in the EPRI reports in detail, as discussed in the Salem, Millstone, and Vogtle SEs. The NRC staff confirmed that the aspects of the transients discussed in those SEs apply equally to this review for Surry, Units 1 and 2. The NRC staff reviewed the discussion of transients in Section 5.2 of EPRI reports 15905, 15906, and 14590, and determined that the transient selection defined in the reports are reasonable for the Surry, Units 1 and 2, plant-specific alternative request because the selection was based on large temperature and pressure variations that are conducive to FCG that is expected to occur in

PWRs. The NRC staff then compared the analysis in the EPRI reports to plant-specific information provided in the licensees submittal.

In Tables A2, A4, A5, and A6, Attachment 2 and Table A7 and A8 of Attachment 3 to the submittal, the licensee evaluated the plant-specific applicability of the transients selected EPRI reports 15905, 15906, and 14590 to the PZR, SG shell, and SG nozzles of Surry, Units 1 and 2.

The NRC staff noted that in some instances the plant-specific transient parameters (e.g.,

temperature and pressure) and the pressurizer insurge/outsurge transient temperature differences were slightly outside the bounds of those transients assessed by the EPRI reports; however, the NRC staff finds that these variances for Surry, Units 1 and 2, are covered by the SS in the EPRI reports.

In the analyses in the EPRI reports there was no separate test conditions included in the transient selection. The licensee stated in Section 5.0 of Attachment 1 to the submittal that pressure tests for Surry, Units 1 and 2, are performed at normal operating conditions and no hydrostatic testing has been performed since the plant began operation. The NRC staff noted that since the pressure tests are performed at normal operating conditions, they are part of Heatup/Cooldown, and therefore test conditions need not be analyzed as a separate transient.

Based on the above, the NRC staff finds that Surry, Units 1 and 2, meets the transient applicability criteria in the EPRI reports. The analyzed transient loads are acceptable for the requested PZR and SG components at Surry, Units 1 and 2. The NRC concludes that the analyses acceptably model transients.

3.2.4.3 Other Operating Loads The NRC staff reviewed the application with regards to weld residual stress and clad residual stress. Weld residual stress and cladding stresses are addressed in EPRI reports 15905, 15906, and 14590. The NRC staff documented the review of these aspects of the EPRI reports in the Salem, Millstone, and Vogtle SEs. The NRC staff determined that no Surry, Units 1 and 2, plant-specific aspects of this submittal warranted additional consideration because of (1) the relatively low sensitivity of the EPRI results on residual stress (Table 8-14 of EPRI report 15905, Table 8-12 of EPRI report 15906, and Table 8-12 of EPRI report 14590) and SS conducted on stress; and (2) the small impact of clad residual stress on the PFM results. Based on this, the NRC staff finds that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on the probability of leakage or rupture beyond the studies documented in the EPRI reports.

Based on the above, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested PZR and SG welds and NIR of Surry, Units 1 and 2. The NRC concludes the analyses acceptably bound other operating loads.

3.2.4.4 Finite Element Analyses The NRC staff reviewed the application with regards to FEA. FEA were conducted in EPRI Reports 3002015905, 3002015906, and 3002014590 as part of the stress analysis portion of the PFM analyses. The NRC staff documented its review in the Salem, Millstone, and Vogtle SEs. The NRC staff determined that no Surry, Units 1 and 2, plant-specific aspects of this application warranted further review because the FEA were performed with the representative component geometries, materials, and loading conditions discussed in Sections 3.2.4.1 and 3.2.4.2 of this SE for which the licensee provided plant-specific information and met the plant-

specific criteria. Based on the above, the NRC staff finds that the plant-specific Surry, Units 1 and 2, submittal is acceptable with regards to FEA.

3.2.5 Fracture Toughness In Attachment 2 to its submittal, the licensee provided information related to the materials of the subject Surry, Units 1 and 2, components. As discussed in Section 3.2.4.1 of this SE, the NRC staff verified that these materials conformed to the requirements of ASME Code,Section XI, Paragraph G-2110. In EPRI reports 15905, 15906, and 14590, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC value of 200 ksiin based on the upper-shelf fracture toughness value in the ASME Code,Section XI, A-4200. The A-4200 fracture toughness curve refers to the same fracture toughness curve in ASME Code,Section XI, Paragraph G-2110. The NRC staff documented the review of the A-4200 fracture toughness value in the Salem, Millstone, and Vogtle SEs. In comparison, the NRC staff determined that the plant-specific Surry, Units 1 and 2, submittal is acceptable with regards to fracture toughness because the materials of the subject Surry, Units 1 and 2, components conform to the requirements of ASME Code,Section XI, Paragraph G-2110.

3.2.6 Flaw Density In Attachment 1 to its submittal, the licensee stated that, per the Vogtle SE, a nozzle flaw density of 0.1 flaws per nozzle should have been used, and that the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10-6 failures per year. Further discussion of this topic as it relates to EPRI Report 3002014590 is contained in the Vogtle SE. The NRC staff noted that the flaw density of 0.1 flaws per nozzle is applicable to a specific plant so long as the component geometries and materials applicability criteria discussed in Section 3.2.4.1 of this SE are met. As discussed in that section of this SE, the licensee provided plant-specific information regarding the geometries and materials of the subject Surry, Units 1 and 2 components and met the applicability criteria.

Based on the above, the NRC staff finds that the appropriate flaw density has been considered, and is, therefore acceptable, for the requested SG NIR of Surry, Units 1 and 2.

3.2.7 FCG Rate The NRC staff reviewed the application with regards to FCG rate. The FCG rate used in EPRI Reports 3002015905, 3002015906, and 3002014590 is based on the ASME Code,Section XI, A-4300 FCG rate. The NRC staff documented its review in detail in the Salem, Millstone, and Salem SEs. The NRC staff noted that FCG rate depends on component material and reactor coolant environment. As discussed in Section 3.2.4.1 of this SE, the licensee provided plant-specific information regarding the materials of the subject Surry, Units 1 and 2, components and met the criteria for component materials. Per the ASME Code,Section XI, the A-4300 FCG rate may be used for light-water-cooled plants. Since Surry, Units 1 and 2, is a PWR, one of the two major types of light-water-cooled reactor designs, the NRC staff determined that the A-4300 FCG rate is appropriate for Surry, Units 1 and 2. Based on the above, the NRC staff finds that the plant-specific Surry, Units 1 and 2, submittal is acceptable with regards to FCG rate.

3.2.8 ISI Schedule and Examination Coverage In Attachment 4 to its submittal, the licensee provided information on the inspection history of the requested PZR welds, and SG welds and NIR of Surry, Units 1 and 2, which consists of the ISI schedule and examination coverage. The licensee stated in Section 5.0 of Attachment 1 to the submittal that for the Surry, Units 1 and 2, PZR preservice inspections (PSI) have been performed followed by ISI examinations over five complete 10-year ISI intervals for Unit 1 and four (4) complete 10-year ISI intervals for Unit 2.

With respect to the Surry Units 1 and 2 SGs:

For the replaced portion of the Surry Unit 1 SGs, PSI examinations have been performed followed by ISI examinations in four complete 10-year ISI intervals following SG replacement. For the portion of the Unit 1 SGs not replaced, PSI examinations have been performed followed by ISI examinations in five complete 10-year ISI intervals.

For the replaced portion of the Surry Unit 2 SGs, PSI examinations have been performed followed by ISI examinations in three complete 10-year ISI intervals following SG replacement. For the portion of the Unit 2 SGs not replaced, PSI examinations have been performed followed by ISI examinations in four complete 10-year ISI intervals.

Given the implementation of ISI in the PFM analyses in the EPRI reports, as the NRC staff explained in the Salem, Millstone, and Vogtle SEs, the NRC staff noted that in terms of PFM modeling ISIs with replacement would be at least as good as ISIs only because replacement is essentially repair of a postulated flaw, while the outcomes of ISI are either repair of a postulated flaw or non-detection and growth of a postulated flaw.

The licensee provided the inspection history of the requested PZR welds, and SG welds and NIR of Surry, Units 1 and 2, in Attachment 4 to the submittal. The licensee also stated that complete examination history from the PSI and the first and second 10-year ISI intervals is not included because many of these older records are stored off-site and were not readily accessible. However, the licensee confirmed that the onsite ISI Program records document that these examinations have been completed in accordance with ASME Code,Section XI requirements. The inspection history also shows that there is no evidence of unacceptable flaws in these components, which is consistent with other known operating history.

Finally, the inspection history shows that some of the examination coverages did not meet the ASME Code,Section XI examination coverage requirement of 90 percent or greater. However, licensees are required to submit a relief request under 10 CFR 50.55a(g)(5)(iii) for ASME Code,Section XI examination requirements that are determined by the licensee to be impractical, which typically includes examination coverages that do not meet the requirement. The NRC staff noted that Attachment 4 to the submittal also provides the reference to the relief requests submitted and approved by the NRC for these instances where the examination coverage requirement was not met.

The NRC staff noted the low examination coverages (50% or less) in some of the requested PZR welds. The NRC staff finds these low examination coverages acceptable in terms of the PFM results because (1) the PZR welds that have low examination coverages do not correspond to the limiting PZR welds analyzed in EPRI report 15905; and (2) the probability of rupture is relatively insensitive to examination coverage for the PZR welds.

During its review, the NRC staff noted that Unit 1 did not complete the required fifth interval ISI examinations of the PZR and SGs at the time of the submittal; thus, it was not able to consider these inspection results during its review of the licenses submittal. By letter dated February 11, 2023, the licensee clarified that all the Unit 1 PZR and SG required examinations have already been completed prior to the end of the fifth 10-year ISI interval, except for the SG C 2-06 weld. The licensee explained that one third of this weld length (368 to 522) was unable to be examined during the last outage of the fifth ISI interval due to scaffolding and schedule issues and that this examination will be performed during the spring 2024 refueling outage. The licensee also confirmed that if indications are detected that exceed the applicable ASME Code,Section XI acceptance standards during the examination of the remainder of the SG C SG 2-06 weld, they will be evaluated in accordance with the ASME Code,Section XI requirements.

The NRC staff has a high level of assurance that the results of this upcoming inspection for the SG C 2-06 weld during the spring 2024 refueling outage will not impact the applicably of the PFM results in the respective EPRI reports for Surry, Unit 1, because only 1/3 of the SG C 2-06 weld is remaining for the licensee to satisfy the fifth 10-year ISI interval requirements, and the past inspection history for the SG Shell to Upper Transition Cone weld, including the latest inspections of the other 2/3 of the SG C 2-06 weld during the 5th ISI interval, had not identified flaws that exceeded the ASME Code Section XI acceptance standards, and the examination coverage for the weld has consistently been greater than 95 percent. Additionally, based on the licensees confirmation that any indications detected during the upcoming inspection that exceed the applicable ASME Code,Section XI acceptance standards, will be evaluated in accordance with the ASME Code,Section XI requirements, ensures that any required successive inspections will be performed in accordance with ASME Section XI IWB-2420, Successive Inspections, irrespective of the licensees proposed alternative in its submittal.

The licensee noted that older examination reports did not report the examination coverage, and therefore, used the more recent examination coverage information to represent the earlier examination coverage that was not documented. This was based on the consistency of the coverages throughout the entire inspection history for the particular weld or component. The NRC staff determined that the licensees use of the more recent examination coverage to be representative of the earlier examination coverage that was not documented to be reasonable because (1) for the SG welds, the examination coverage values are relatively high (greater than 90 percent); and (2) for the PZR welds, probability of rupture is relatively insensitive to examination coverage.

Based on this discussion, the NRC staff finds the Surry, Units 1 and 2, inspection history of the subject PZR welds, and SG welds to be acceptable. Based on the above and the inspection history of the PZR welds, and SG welds and NIR of Surry, Units 1 and 2, the NRC staff finds that the PFM approach of EPRI reports 15905, 15906, and 14590 sufficiently represent the requested components for Surry, Units 1 and 2, with respect to ISI schedule and examination coverage.

3.2.9 Other Considerations The NRC staff reviewed the application concerning initial flaw depth and length distribution, probability of detection, models, uncertainty, and convergence. The NRC staff noted that these other considerations of the analyses in the EPRI reports do not depend on plant-specific information, as compared to component geometries, materials, and transient selection for which

the licensee provided plant-specific information to ensure applicability of the analyses in the reports, as discussed previously.

Initial flaw depth and length distribution do not depend on plant-specific information because the flaw distribution used was based on fabrication flaws instead of service-induced flaws.

Probability of detection, which is associated with volumetric examinations, does not depend on plant-specific information because the corresponding components in different plants are subject to the same volumetric examination requirements of the ASME Code,Section XI. The models (i.e., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in fracture mechanics analyses. Uncertainty and convergence do not depend on plant-specific information because these are part of the overall PFM analyses that were addressed in the sensitivity studies and sensitivity analyses in the EPRI reports.

The NRC staff previously reviewed the applicable aspects of these considerations as used in EPRI Reports 3002015905, 3002015906, and 3002014590, and documented their acceptability in detail in the Salem SE, Millstone SE, and Vogtle SE. Since these considerations are not dependent on plant-specific information, the NRC staff finds that the plant-specific Surry, Units 1 and 2, submittal is acceptable in terms of these considerations.

3.2.10 PFM Results Relevant to Proposed Alternative In Section 5.0 of Attachment 1 to the submittal, the licensee stated that based on the PFM results, after PSI, no other inspections are required for up to 80 years of plant operation to meet the acceptance criterion of 1x10-6 failures per year. Similar statements are made in EPRI reports 15905, 15906, and 14590. The NRC staff does not find this conclusion acceptable since it does not account for the effect of the combination of the most significant parameters or the added uncertainty of low probability events. More significantly, the NRC staff considers this conclusion to be a solely risk-based approach, which is inconsistent with NRC policy that calls for risk insights to be considered together with other factors rather than sole reliance on risk-based approaches. Post fabrication examinations are critical in supporting necessary performance monitoring goals including monitoring and trending; bounding uncertainties; validating/confirming analytical results; and providing timely means to identify novel and/or unexpected degradation.

Notwithstanding the discussion above, the PFM analyses in the EPRI reports investigated several ISI examination schedule scenarios, which include PSI followed by various ISI examinations. The PFM results relevant to the proposed alternative for Surry, Units 1 and 2, are those resulting from an ISI schedule scenario that closely matches that which is discussed in Section 3.2.8 of this SE. The relevant PFM results show that the probability of rupture is below the acceptance criterion of 1x10-6 failures per year. Based on the above and the discussions in Sections 3.2.1 through 3.2.9 of this SE, the NRC staff finds that the proposed alternative for Surry, Units 1 and 2, for the requested PZR welds, and SG welds and NIR of Surry, Units 1 and 2, would result in a PoF per year that is below the acceptance criterion of 1x10-6 failures per year.

3.2.11 Performance Monitoring

Background

Performance monitoring, such as ISI programs, is a necessary component such as described by the NRC five principles of risk-informed decision making. Analyses, such as PFM, work along with performance monitoring to provide a mutually supporting and diverse basis for facility condition and maintenance that is within its current licensing basis. An adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation of continued appropriateness of associated analyses, and a timely method to detect novel/unexpected degradation. These characteristics were presented, for example, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277; agenda and slides, respectively). Previously, the NRC staff has applied binomial statistics and Monte Carlo methods to augment evaluation of periods beyond 20 years. The methods used by the NRC staff were presented at a May 25, 2022, public meeting (ML22144A345, and ML22143A840, meeting notice and presentation respectively).

Surry, Units 1 and 2, Evaluation The proposed alternative for Surry, Units 1 and 2 would result in a significant amount of time before another examination was performed on several welds under the submittal. Specifically, the components that have not yet received an ISI examination for the fifth 10-year ISI interval.

The NRC staff requested the licensee to provide a performance monitoring plan for these components which extend substantially beyond 20 years between examinations (see response to RAI-1 in the February 11, 2023 supplement). The 20-year threshold is consistent with prior precedent where U.S. licensees have sought examination relief from prescriptive ASME Section XI requirements. For example, the NRC has conditioned Code Case N-864 on reactor vessel threads in flange examinations to require U.S. licensees to perform ultrasonic examinations at least every third inservice inspection interval in Regulatory Guide 1.147, Revision 20 (ML21181A222).

In its RAI response, the licensee explained that performance monitoring for all applicable welds/components will resume with the start of the seventh 10-year ISI Interval beginning May 10, 2034, in accordance with its ASME Code Section XI, 10-year ISI Interval plan.

Additionally, by letter dated March 2, 2023, the licensee confirmed that it will resume the ASME Section XI Code required weld inspections during the first period of the seventh 10-Year ISI Interval as follows:

at least one Surry, Unit 1 or Unit 2, steam generator will be inspected during the first period of the 7th ISI interval and at least one Surry, Unit 1 or Unit 2, pressurizer will be inspected during the first period of the seventh ISI interval.

The NRC staff noted that the seventh 10-year ISI interval ends October 13, 2043, and May 9, 2044, for Surry, Units 1 and 2, respectively.

In its submittal the licensee also discussed system leakage tests as providing assurance of safety for the proposed alternative. However, the NRC staff noted that the visual examinations performed during system leakage tests may not directly detect the presence or extent of degradation; may not provide direct detection of aging effects prior to potential loss of structure or intended function; and do not provide sufficient validating data necessary to confirm the

modeling of degradation behavior in the subject PZR and SG welds. The NRC staff noted that leakage tests provide complementary additional performance monitoring to the ISI examinations but would not, in isolation, be sufficient.

Prolonged periods of plant operation without inspection may result in a lack of monitoring and trending capacity and does not provide a sufficient basis for continued adequacy of component integrity. Consequently, the NRC staff performed a variety of simulations regarding potential inspection scenarios and the likelihood that such proposals would support the necessary characteristics of adequate performance monitoring. The NRC staff sought to understand the capacity of the proposed performance monitoring plan to detect potential novel degradation.

These simulations were conducted using binomial statistics and Monte Carlo methods. Based on these simulations, which encapsulates the Surry, Units 1 and 2 conditions, the NRC staff determined that the previously conducted and proposed volumetric inspections during the first period of the seventh ISI interval would constitute sufficient performance monitoring in concert with the other aspects of the submittal reviewed by the NRC staff. The NRC staff noted that some supporting monitoring and trending information for these components will continue to be accrued at other facilities, naturally spread by date of application, interval schedules, and other factors, providing further assurance that adequate monitoring and trending will continue.

Based on the above, and given the supplemental information in the RAI response, the NRC staff determined that inspections for the subject components could be deferred during the proposed period because an adequate level of performance monitoring is maintained for the components.

The NRC noted that, as confirmed by the licensee in its submittal and supplement letters dated February 11, 2023, and March 2, 2023, the required ASME Code,Section XI inspections will resume in the first period of the seventh ISI interval.

3.9 CONCLUSION

As set forth above, the NRC staff determined that the licensees proposed alternative as discussed above, for the requested components, provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative for Surry, Unit 1, for the duration of the sixth 10-year ISI interval, and for Surry, Unit 2, for the duration of the fifth 10-year ISI interval and through the sixth 10-year ISI interval.

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, NRR O. Yee, NRR

ML23073A191 *Via SE Input OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC* NRR/DORL/LPL2-1/BC NAME JKlos KGoldstein ABuford MMarkley DATE 3/09/2023 4/20/2023 3/9/2023 4/21/2023 OFFICE NRR/DORL/LPL2-1/PM NAME JKlos DATE 4/21/2023