ML24086A541

Authorization of Request for Alternative ANO1-ISI-037 Regarding Extension of Reactor Vessel Inservice Inspection Interval

ML24086A541
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/10/2024
From:	Jennivine Rankin NRC/NRR/DORL/LPL4
To:	Entergy Operations
Wengert T, NRR/DORL/LPL4
References
EPID L-2023-LLR-0028
Download: ML24086A541 (1)

Text

April 10, 2024

ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.

N-TSB-58 1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 1 - AUTHORIZATION OF REQUEST FOR ALTERNATIVE ANO1-ISI-037 REGA RDING EXTENSION OF REACTOR VESSEL INSERVICE INSPECTION INTERVAL (EPID L-2023-LLR-0028)

Dear Site Vice President:

By letter dated June 8, 2023 (Agencywide Documents Access and Management System Accession No. ML23159A269), as supplemented by letter dated December 14, 2023 (ML23348A384), Entergy Operations, Inc. (the licensee), submitted request for alternative ANO1-ISI-037 to the U.S. Nuclear Regulatory Commission (NRC).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in Relief Request No. ANO1-ISI-037 for the remainder of the fifth and sixth10-year inservice inspection (ISI) intervals at Arkansas Nuclear One, Unit 1 (ANO-1) for American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, table IWB-2500-1, Examination Categories B-A and B-D, on the basis that the alternative provides an acceptable level of quality and safety. The item numbers in scope of Examination Categories B-A and B-D are in the reactor vessel (RV) and RV welded nozzles of ANO-1.

As set forth in the enclosed safety evaluation, the NRC staff has determined that the licensees proposed alternative would provide an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI requirements and inspection items specified and referenced in Relief Request No. ANO1-ISI-037. Accordingly, the 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 Relief Request No.

ANO1-ISI-037 at ANO-1 for the fifth and sixth 10-year ISI intervals.

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

If you have any questions, please contact the ANO-1 Project Manager, Thomas Wengert at (301) 415-4037 or by email at Thomas.Wengert@nrc.gov.

Sincerely,

Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket No. 50-313

Enclosure:

Safety Evaluation

cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

REQUEST FOR ALTERNATIVE ANO1-ISI-037

REGARDING EXTENSION OF REACTOR VESSEL INSERVICE INSPECTION INTERVAL

ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1

DOCKET NO. 50-313

1.0 INTRODUCTION

By letter dated June 8, 2023 (Agencywide Documents Access and Management System Accession No. ML23159A269), as supplemented by letter dated December 14, 2023 (ML23348A384), Entergy Operations, Inc. (Entergy, the licensee), submitted request for alternative ANO1-ISI-037 to the U.S. Nuclear Regulatory Commission (NRC).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in ANO1-ISI-037 for the fifth and sixth 10-year inservice inspection (ISI) intervals at Arkansas Nuclear One, Unit 1 (ANO-1) for American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI, Inservice Inspection of Nuclear Power Plant Components, table IWB-2500-1, Examination Categories B-A and B-D, on the basis that the alternative provides an acceptable level of quality and safety. The item numbers in scope of Examination Categories B-A and B-D are in the reactor vessel (RV) and RV welded nozzles at ANO-1.

2.0 REGULATORY EVALUATION

Regulatory Requirements

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI.

The regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, states, in part, that alternatives to the requirements of 10 CFR 50.55a(b) through (h) may be used, when authorized by the Director, Office of Nuclear Reactor Regulation, if (1) the proposed alternatives 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.

Enclosure

The regulation in 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock events, requires that the reference temperature of the RV materials be within specific values to protect reactor pressure vessel (RPV) materials against pressurized thermal shock (PTS) events.

The regulation in 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events, specifies alternate rules for protection against PTS events.

Regulatory Guidance

Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ML003740284), specifies guidance on determination of embrittlement shift of RV materials due to irradiation.

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, dated November 2002 (ML023240437), specifies guidance on assessment of risk-informed decisions on changes to the plant-specific current licensing basis.

NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), dated March 2010 (ML15222A848), specifies recommended PTS screening limits for RV materials.

3.0 TECHNICAL EVALUATION

3.1 ASME Code Components Affected

The affected components are the ANO-1 RV shell welds and nozzles associated with the following ASME Code,Section XI examination categories and item numbers. These examination categories and item numbers are from subarticle IWB-2500 and table IWB-2500-1 of the ASME Code,Section XI.

Examination Category Item No. Description B-A B1.10 Shell Welds B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.20 Head Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.50 Repair Welds B-A B1.51 Beltline Welds B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section

Note: Examination Category B-A welds are defined as Pressure Retaining Welds in Reactor Vessel. Examination Category B-D welds are defined as Full Penetration Welded Nozzles in Vessels.

3.2 Applicable ASME Code Edition and Addenda

The fifth 10-year ISI interval for ANO-1 is scheduled to end on May 30, 2027. The Code of record for the fifth 10-year ISI interval is ASME Code,Section XI, 2007 Edition through 2008 Addenda. The applicable Code for the sixth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a, Codes and standards.

3.3 Applicable ASME Code Requirement

Paragraph IWB-2411, Inspection Program, requires volumetric examination of essentially 100 percent of RV pressure-retaining welds identified in table IWB-2500-1 once each 10-year ISI interval.

3.4 Reason for Request

The licensee is requesting an alternative from the IWB-2411 Inspection Program that requires volumetric examination of essentially 100 per cent of RV pressure-retaining Examination Categories B-A and B-D welds once each 10-year ISI interval. The licensee stated that extension of the interval between examinations of Categories B-A and B-D welds from 10 years to up to 20 years will result in a reduction in radiation exposure and examination costs.

3.5 Proposed Alternative and Basis for Use

In section 5, Proposed Alternative and Basis for Use, of enclosure 2 to the submittal dated June 8, 2023, the licensee proposed to not perform the ASME Code required volumetric examination of the ANO-1 RV full penetration pressure-retaining Examination Categories B-A and B-D welds for the fifth 10-year ISI interval currently scheduled for 2027. Instead, the licensee proposes to perform these volumetric examinations in 2036. In its response to Request for Additional Information (RAI)-1 in the supplemental letter dated December 14, 2023, the licensee revised the proposed rescheduled inspection date of 2036 to prior to May 20, 2034, during the sixth ISI interval. The expiration date for the Renewed Facility Operating License No. DPR-51 for ANO-1 (ML053130314) is May 20, 2034. The licensee stated that the proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238, Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval. PA-MSC-0120 (ML11153A033), since the implementation plan originally showed the next inspection scheduled for 2028 for ANO-1. This implementation plan is used in conjunction with the use of the methodology in WCAP-16168-NP-A, Risk-Informed Extension of the Reactor Vessel Inservice Inspection Interval, Revision 3 (ML11306A084), hereafter referred to as WCAP-A in this safety evaluation (SE). The licensee further stated that this proposed inspection schedule is considered to have a minor impact on the future inspection plan and distribution of inspections over time.

In accordance with 10 CFR 50.55a(z)(1), the licensee proposed an alternate (i.e., extended) ISI interval on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in RG 1.174. The licensee stated that the methodology used to conduct this analysis is based on that defined in WCAP-A. The licensee stated that the results of the calculations for ANO-1 were compared to those obtained from the Babcock and Wilcox (B&W) pilot plant evaluated in WCAP-A. The licensee stated that the parameters for ANO-1 are bounded by the results of the B&W pilot plant evaluation.

3.6 Duration of Proposed Alternative

The licensee stated that the requested alternative is applicable to the ANO-1 ISI program for the fifth and sixth 10-year ISI intervals.

3.7 NRC Staff Evaluation

The licensees proposed extended ISI interval in ANO1-ISI-037 is based on a risk-informed RV fracture mechanics analysis that was performed in accordance with the NRC staff-approved, risk-informed flaw analysis methods in WCAP-A. The methodology in WCAP-A was developed by the Pressurized Water Reactor Owners Group (PWROG) to satisfy through-wall cracking frequency (TWCF) criteria, specifically the 95 th percentile total TWCF (TWCF 95-TOTAL), for pressurized water reactors (PWRs) estab lished in NUREG-1874 and the delta large early release frequency (LERF) criteria specified in RG 1.174.

In section 3.4 of the NRC staffs SE dated July 26, 2011 (ML111610242), for WCAP-A (hereafter referred as WCAP-A SE), the staff specified plant-specific information that licensees must submit for alternative requests that are based on the methodology of WCAP-A. Tables A-1, A-2, and A-3 in appendix A of WCAP-A show the format used for providing the plant-specific information. The licensee provided the plant-specific information for ANO-1 in tables 1, 2, and 3 of section 5 of enclosure 2 to the June 8, 2023, submittal, as updated in enclosure 3 to the supplement. The staff evaluated the plant-specific information in the following subsections.

3.7.1 Identification of Limiting Design Basis Transients and Cladding Layers

Regarding the PTS transients, the licensee identified in table 1 of enclosure 2 to the June 8, 2023, submittal, that the transients are defined in NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants (ML042880482),

hereafter referred as PTS Generalization Study, and that those transients serve as the limiting design-basis transients for the RV welds that were assessed in ANO1-ISI-037. The NRC staff verified that for B&W-designed PWRs, such as ANO-1, the PWROGs methodology in WCAP-A uses the PTS transients that were defined in NUREG-1874 and clarified in the PTS Generalization Study as the limiting PTS transi ents for the PWROGs risk-informed flaw analysis that was included in WCAP-A. Therefore, the staff finds the licensees transient basis to be acceptable based on the information in NUREG-1874 and the PTS Generalization Study, and the staffs conclusions in the WCAP-A SE that the PTS transient characteristics for a given nuclear steam supply system design of U.S. PWR light water reactor facilities are generically applicable for all PWRs designed by the same reac tor nuclear steam supply system vendor (i.e.,

B&W for ANO-1).

Regarding the cladding layers, the licensee reported in table 1 of enclosure 2 to the June 8, 2023, submittal that the cladding for the RV at ANO-1 was deposited using a single layer. The NRC staff confirmed that the RV cladding at the B&W plant analyzed in WCAP-A was deposited using a single layer. Thus, for the proposed alternative, the staff concludes that the licensee did not need to evaluate the impacts that multiple pass layers would have on the design of the RV cladding at ANO-1 because: (1) the cladding layer at ANO-1 was deposited as a single layer, and (2) the design of the cladding layer at ANO-1 is consistent with and bounded by the staffs evaluation of the cladding layer in the WCAP-A SE.

3.7.2 Frequency and Severity of Design Basis Transients

In section 3.4 of the WCAP-A SE, the NRC staff stated, in part, that:

Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all of its design basis transients; and (b) identify the design bases transients that contribute to significant fatigue crack growth.

In table 1 of enclosure 2 to the June 8, 2023, submittal, the licensee indicated that the plant-specific basis for frequency and severity of design-basis transients is bounded by the 12 cycles/reactor year of plant heatup and cooldown transients of the associated pilot plant assessed in WCAP-A. The licensee provided in enclosure 3 to the submittal (Framatome Document No. 86-9352400-000, Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location, Revision 0) an equivalent 20-year fatigue crack growth (FCG) analysis for ANO-1 using plant-specific transient s, including heatup and cooldown. The licensee provided this equivalent FCG analysis to address the NRC staffs expectations in the WCAP-A SE. The licensees equivalent FCG analysis listed the design-basis transients for ANO-1 that contribute to FCG and showed that the 20-year FCG value due to ANO-1 plant-specific transients is less than that due to 12 heatup and cooldown cycles per year. The equivalent FCG analysis also indicated that the highest contributor to crack propagation is from the power change transient, power loading transient, and heatup/cooldown transients, all of which are included in the 20-year FCG analysis for ANO-1.

The NRC staff confirmed in the NRC staff-approved WCAP-A methodology that the PWROG established that the reactor coolant system heatup and cooldown transients are the limiting design transients for RPV fatigue flaw growth in B&W-designed PWRs. The NRC staff also confirmed that the PWROG established 12 cycles/reactor year as the maximum bounding number of heatup and cooldowns that could occur for B&W-designed PWRs. The staff confirmed that the design-basis transients used as input to the licensees equivalent FCG analysis for ANO-1 are consistent with design transients specified in Amendment 31 of the ANO-1 Safety Analysis Report (ML23180A110). The staff also noted that the ANO-1 equivalent FCG value for 12 heatup and cooldown cycles per year was less than the corresponding FCG value for Oconee Nuclear Station, Units 1, 2, and 3 (Oconee), which was the B&W pilot plant analyzed in WCAP-A. The FCG value for Oconee was provided as part of an alternative request for Oconee submitted by letter dated January 19, 2021 (ML21019A276), which was subsequently authorized by the NRC staff by letter dated November 19, 2021 (ML21281A141).

Based on the discussion above, the NRC staff finds that the licensees equivalent FCG analysis for ANO-1 is acceptable because it is sufficiently bounded by the FCG analysis analyzed and established for B&W-designed PWR units in WCAP-A.

3.7.3 Scope and Schedule for Inspecting the RV Welds During the 20-Year ISI Interval

In section 3.4 of the WCAP-A SE, the NRC staff stated that licensees should identify the ISI schedule for RV weld examinations that will be performed during the proposed 20-year ISI interval. The WCAP-A SE also established the staffs position that the dates for the weld inspections must be within one refueling cycle of the revised dates identified for inspection in the implementation plan in PWROG Letter No. OG-10-238 (ML11153A033).

In section 5 and table 2 of enclosure 2 to the June 8, 2023, submittal, as supplemented, the licensee proposes not to perform required ASME Code volumetric examinations of the applicable RV weld components of ANO-1 for the fifth 10-year ISI interval currently scheduled for 2027. The licensee proposes instead to perform these volumetric examinations prior to May 20, 2034. The licensee stated that this proposed inspection date is a deviation from the implementation plan in PWROG Letter No. OG-10-238 since the plan reflects the next inspection for ANO-1 to be performed in 2028. The licensee stated that the resulting impact to the implementation plan in PWROG Letter No. OG-10-238 would increase the number of inspections [as specified in the plan] in 2034 (from five to six) and decrease the number of inspections in 2028 (from five to four). The licensee also stated that, based on Figure 3 and Figure 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

Based on the discussion above, the NRC staff finds that the licensees proposed scope and schedule for inspecting the subject RV welds are acceptable because the deviation from implementation plan in PWROG Letter No. OG-10-238 is only a minor deviation, and therefore, would have little impact on the stream of performance monitoring data specified in PWROG Letter No. OG-10-238.

3.7.4 Relevant Operating Experience - Summary of ISI Results

In section 3.4 of the WCAP-A SE, the NRC sta ff established its position that a licensee submitting risk-informed ISI extension(s) for its RVs should report the results of its prior ISI inspections of the applicable RV weld locations.

In table 2 of enclosure 2 to the June 8, 2023, submittal, the licensee identified that it performed preservice inspection and four 10-year ISIs. The licensee reported that one subsurface indication was identified during the fourth 10-year ISI located in the lower shell longitudinal weld.

The licensee stated that the indication is acceptable per table IWB-3510-1 of Section XI of the ASME Code and that the indication is not within the inner 1/10 th or 1 inch of the RV thickness.

Based on the discussion above, the NRC staff finds the licensees ISI results for the subject RV welds acceptable because the one subsurface indication that was detected was acceptable per ASME Code,Section XI, and outside the depth within the RV thickness where additional attention is needed, as specified in 10 CFR 50.61a.

3.7.5 Susceptibility to Underclad Cracking of RV Forgings

In section 3.4 of the WCAP-A SE, the NRC staff determined that licensees with RVs containing forgings that are susceptible to underclad cracking and have RT MAX-FO values exceeding 240 degrees Fahrenheit (°F) (699.67 degrees Rankine (ºR)) must submit a plant-specific evaluation because the analyses performed in the WCAP-A are not applicable (i.e., the scope of analyses in WCAP-A do not cover the evaluation of RV underclad cracks for forgings with high RTMAX-FO values).

In table 3 of enclosure 2 to the June 8, 2023, submittal, as supplemented, the licensee reported a maximum (limiting) RTMAX-FO value of 508.28ºR at 54 effective full power years (EFPY), which is less than 699.67ºR and bounded by the value provided in Framatome Document No. 86-9352400-000 (enclosure 3 to the June 8, 2023, submittal). The NRC staff also noted that the licensee appropriately addressed RV underclad cracking in the license renewal application for ANO-1, and that the staff found acceptable the licensees underclad cracking time-limited

aging analysis (TLAA) and aging management bases for RV forging components in section 4.8.1, Reactor Vessel Underclad Crac king, of NUREG-1743, Safety Evaluation Report Related to License Renewal of Arkansas Nuclear One, Unit 1, dated May 2001 (ML011640099, ML011640177, and ML011640217).

Based on the discussion above, the NRC staff concludes that the licensee did not need to include any supplemental plant-specific evaluations for ANO-1 in its submittal (i.e., beyond the unit-specific flaw distribution, RT MAX-XX, limiting TWCF95-XX, and TWCF95-TOTAL included for ANO-1 in the submittal) because: (1) the licensee has provided sufficient demonstration that ANO-1 does not have RV forgings with RT MAX-FO values greater than 240°F (699.67ºR); and (2) the licensee has appropriately managed underclad cracking in the RV forgings through implementation of the applicable TLAA that the staff approved in NUREG-1743.

3.7.6 Submittal of Information Required by 10 CFR 50.61a(e)

In section 3.4 of the WCAP-A SE, the NRC staff stated, in part, that [l]icensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61a [10 CFR 50.61a(e)]. In its response to RAI-2 in the supplemental letter dated December 14, 2023, the licensee confirmed that alternative request ANO1-ISI-037 is the first interval extension request for ANO-1 that is based on the WCAP-A methodology.

Therefore, the staff determined that the licensee does not need to include the information required by 10 CFR 50.61a(e), Examination and flaw assessment requirements, for alternative request ANO1-ISI-037.

3.7.7 TWCF Evaluation

In section 3.4 of the WCAP-A SE, the NRC sta ff established its position that the maximum adjusted reference temperatures and 30 foot-pounds Charpy V-notch energy shifts in reference temperature values (i.e., RTMAX-X and T30 values, as defined in 10 CFR 50.61a) may be calculated using the methods documented in RG 1.99, Revision 2, or in an alternate NRC-approved methodology using these types of parameters. The WCAP-A SE also stated that licensees submittals should include the material property and fluence information related to these parameters and that appendix A of WCAP-A (table A-3) identifies the information needed to be submitted.

In table 3 of enclosure 2 to the June 8, 2023,submittal, the licensee included the material property and neutron fluence data, T30 values, and RTMAX-XX values for the RV base metal and weld components of ANO-1 at 54 EFPY. The licensee stated that material properties are based on ANP-3300, Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY, dated June 2014 (ML14241A241), and that fluence projections and material properties not included in ANP-3300 were provided by Framatome. In the supplement dated December 14, 2023 (response to RAI-3), the licensee clarified that the fluence projection values have been updated to the those in ANP-3300, Revision 1 (ML14330A250) to be consistent with the current licensing basis (CLB) values. In its response to RAI-4 in its supplement, the licensee clarified that ANP-3300, Revision 1 is the CLB analysis of record for ANO-1 and provided updated analysis values in table 3 of enclosure 3 to the supplement.

The licensee calculated a TWCF 95-TOTAL value of 1.08x10-10 per year for the RV, as shown in the table 3 of enclosure 3 of the supplement. This TWCF 95-TOTAL value is less than the limiting TWCF95-TOTAL value of 4.42x10-7 events per year approved in WCAP-A for B&W plants. The NRC staff verified the calculations in the TWCF 95-XX column of table 3 of enclosure 3 of the

supplement, and noted that the licensees TWCF 95-XX values were conservative compared to the staffs calculations. Therefore, the staff finds that the licensees TWCF 95-TOTAL value for the ANO-1 RV is acceptable.

The NRC staff noted that the methodology in WCAP-A conservatively sets the TWCF 95-TOTAL equal to the LERF value for the RV that may result from initiation of the postulated, limiting PTS event at a plant. Thus, based on the staffs independent calculations and verifications discussed above, the staff determined that the TWCF 95-TOTAL value for the ANO-1 RV meets the limit of 1x10-7 early release events per reactor year that is established for LERF values in RG 1.174.

3.7.8 NRC Staff Evaluation Summary

Based on the evaluations in sections 3.7.1 through 3.7.7 of this SE, the NRC staff determined that the licensee has satisfied all plant-specific information items specified in the WCAP-A SE.

Therefore, the staff finds that the licensees proposed risk-informed alternative provides an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI requirements and inspection items specified and referenced in ANO1-ISI-037.

4.0 CONCLUSION

As set forth above, the NRC staff has determined that the proposed alternative in the licensees request would provide an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI requirements and inspection items specified and referenced in Relief Request No. ANO1-ISI-037. Accordingly, the 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 a proposed alternative in Relief Request No. ANO1-ISI-037 at ANO-1 for the fifth and sixth 10-year ISI intervals.

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

Principal Contributor: D. Dijamco, NRR

Date: April 10, 2024

ML24086A541 *by email NRR-028 OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DNRL/NPHP/BC* NRR/DORL/LPL4/BC*

NAME TWengert PBlechman ABuford JRankin DATE 4/5/2024 4/4/2024 2/23/2024 4/10/2024