ML23142A106

Authorization and Safety Evaluation for Alternative Request No. ANO1-ISI-035

ML23142A106
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/24/2023
From:	Jennifer Dixon-Herrity Plant Licensing Branch IV
To:	Entergy Operations
Wengert T
References
EPID L-2022-LLR-0050
Download: ML23142A106 (1)

Text

May 24, 2023 ARKANSAS NUCLEAR ONE, UNIT NO. 1 - AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NO. ANO1-ISI-035 (EPID L-2022-LLR-0050)

LICENSEE INFORMATION Recipients Name and Address: ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.

N-TSB-58 1448 S.R. 333 Russellville, AR 72802 Licensee: Entergy Operations Inc.

Plant Names and Units: Arkansas Nuclear One, Unit 1 (ANO-1)

Docket Nos.: 50-313 APPLICATION INFORMATION Submittal Date: May 31, 2022 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML22151A322 Supplement Dates: January 30, 2023 Supplement ADAMS Accession Nos.: ML23030B329 Applicable Inservice Inspection (ISI) Interval and Interval Start/End Dates: ANO-1 fifth 10-year ISI interval start date is May 31, 2017, and is scheduled to end on May 30, 2027.

ANO-1 sixth 10-Year ISI interval start date is May 31, 2027, and is scheduled to end on May 20, 2034.

Alternative Provision: The licensee requested an alternative from certain requirements under Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2) on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

ISI Requirement: Pursuant to 10 CFR 50.55a(g)(6)(ii), the U.S. Nuclear Regulatory Commission (NRC) may require the licensee to follow an augmented ISI program for systems and components for which the NRC deems that added assurance of structural reliability is necessary.

The regulations in 10 CFR 50.55a(g)(6)(ii)(D), Augmented ISI requirements: Reactor vessel head inspections, require licensees of pressurized water reactors (PWRs) to augment their ISI of the reactor vessel head with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code (ASME Code) Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, with conditions.

Furthermore, the regulation, 10 CFR 50.55a(g)(6)(ii)(D)(1), Implementation, states:

Holders of operating licenses or combined licenses for pressurized-water reactors as of or after June 3, 2020, shall implement the requirements of ASME BPV Code Case N-729-6 instead of ASME BPV Code Case N-729-4, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (8) of this section, by no later than one year after June 3, 2020. All previous NRC-approved alternatives from the requirements of paragraph (g)(6)(ii)(D) of this section remain valid.

Table 1 of ASME Code Case N-729-6 requires a volumetric or surface examination per Item B4.40 with the following scope and frequency:

All nozzles, not to exceed two inspection intervals (nominally 20 calendar years).

Applicable Code Edition and Addenda

ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 1, 2007 Edition through 2008 Addenda.

The regulation in 10 CFR 50.55a(g)(6)(ii)(D) mandates the use of ASME Code,Section XI, Division 1, Code Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, with certain conditions.

Brief Description of the Proposed Alternative: The licensee is proposing an alternative to defer the volumetric or surface examination of reactor vessel closure head (RVCH) Penetration Number 0-1 (Penetration No. 1) in ANO-1 to the end of the current, renewed ANO-1 operating license (May 20, 2034).

The licensee is requesting an alternative in accordance with 10 CFR 50.55a(z)(2), due to the hardship of compliance with the regulation without a compensating increase in the level of quality and safety related to the removal and reinstallation of the post-accident reactor vessel water level (RADCAL) instrument installed in Penetration No. 1.

For additional details on the licensees request, refer to the documents located at the ADAMS accession numbers identified above.

STAFF EVALUATION The NRC staff reviewed and evaluated the licensees request pursuant to 10 CFR 50.55a(z)(2),

in that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee explained that the hardship associated with performance of the volumetric or surface examination in accordance with ASME Code Case N-729-6 for Penetration No. 1 is that such an examination would require removal and subsequent reinstallation of the RADCAL instrument, which would entail significant worker dose exposure and considerable risks. The licensee further explained that without removing the instrument, the tight gap between the RADCAL instrument guide assembly and the inside of Penetration No. 1 precludes ultrasonic testing or eddy current testing using a blade probe. Removal of the RADCAL instrument, including its integral guide assembly, would require unbolting the flanged joint connecting the instrument to the primary pressure boundary just above the RVCH.

This bolted joint is intended to be a permanent mechanical connection. The licensee stated that the risks of removal and reinstallation of the RADCAL instrument could include unintended damage to the instrument and other components, as well as the potential for leakage of reactor coolant at the flanged joint, which could cause corrosion of the upper RVCH surface.

The NRC staff reviewed the licensees hardship basis, as well as its documented options to address these issues, and finds that the licensee took all reasonable steps to evaluate possible alternatives and that the licensee provided an adequate basis to justify the hardship.

Therefore, the NRC staff finds that the licensee meets the hardship requirement of 10 CFR 50.55a(z)(2).

The NRC staff notes that ANO-1 has a replacement RVCH, which includes Alloy 690 materials that are more resistant to primary water stress corrosion cracking (PWSCC) than the original RVCH materials. In accordance with 10 CFR 50.55a(g)(6)(ii)(D), the NRC mandated the use of ASME Code Case N-729-6, which was published in the Federal Register on May 4, 2020 (85 FR 26540). As part of this action, the NRC staff extended the volumetric and surface examination frequency of nozzles and partial penetration welds of PWSCC-resistant materials, such as Alloy 690, in the RVCH from every 10 years to every 20 years. This change was based, in part, on the NRC staffs review of Materials Reliability Program [MRP]: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375) (ML14283A046), and the proposal of a factor of improvement (FOI) of 12 when using these materials in replaced RVCHs, as summarized in the Electric Power Research Institute publicly available slides dated August 21, 2014 (ML14237A256). The MRP-375 analysis was a generic bounding analysis for all conditions. The licensee supported its proposed alternative by performing a hypothetical flaw analysis based on the plant-specific conditions for Penetration No. 1. The licensee concluded that a FOI of 8.2 was required for the limiting hypothetical flaw analysis case.

The NRC staff reviewed the licensees hypothetical flaw analysis provided as an attachment to its submittal. The NRC staff found the input requirements used by the licensee were consistent with similar RVCH analyses. Specifically, the NRC staff found the crack growth rate and hypothetical flaw paths were acceptable. The NRC staff determined that the RVCH temperature was conservative and consistent with similar previously authorized alternatives. Additionally, the NRC staff found the FOI approach to be reasonable, given the previous work with MRP-375 and previous alternatives the NRC staff authorized for other licensees to address RVCH examination frequency extensions. The NRC staff reviewed the licensees weld residual stress profiles relative to other alternatives and Appendix A of Materials Reliability Program Generic Evaluation of Examination Coverage Requirements for Reactor Pressure Vessel Head Penetration Nozzles, Revision 1 (MRP-95R1) (ML043200602), and found that the weld residual stress profiles were consistent with other 0-degree centerline nozzle penetrations, with a relatively lower weld residual stress maximum value, granting slower PWSCC growth for this

specific penetration nozzle configuration. Finally, the NRC staff found the licensees hypothetical flaw analysis model basis was consistent with the development of the re-inspection frequency methodology of ASME Code Case N-729-6. Therefore, the NRC staff found the licensees hypothetical flaw analysis methodology acceptable.

The NRC staff performed a review of the licensees results to evaluate the potential uncertainties associated with PWSCC flaw analysis. The FOI of 8.2, determined by the licensee, provides a margin of approximately 32 percent compared to the NRC-approved FOI for the current regulatory condition. Given this margin and the acceptable methodology, the NRC staff finds that the licensees hypothetical flaw analysis provides reasonable assurance of structural integrity of the subject penetration nozzle for the proposed alternatives period of extended examination frequency.

The NRC staff also considered the licensees additional factors to support the technical basis of the proposed alternative. These included the following items:

1. Operating experience showed no PWSCC in Alloy 690 materials being identified in the fleet, as well as the previous inspections of all other Alloy 690 penetration nozzles in the same RVCH at ANO-1.

2. The licensee would continue to maintain the frequency of the required bare metal visual examination of every 5 years for the RVCH.

3. The licensee has incorporated enhanced leakage detection capability at ANO-1 consistent with the guidance of WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS [Reactor Coolant System] Leakage Action Levels and Response Guidelines for Pressurized Water Reactors (ML070310082).

The NRC staff finds that each of these items provides additional defense-in-depth in supporting the hypothetical flaw analysis calculations to support the conclusion that structural integrity of the RVCH at ANO-1 will be maintained throughout the duration of the licensees proposed alternative.

Given the licensees technical basis to provide reasonable assurance of structural integrity and its identified hardship, the NRC staff finds that the licensee has provided an adequate technical basis to support deferral of the volumetric or surface examination of Penetration No. 1 in the ANO-1 RVCH to the end of the current, renewed ANO-1 operating license (May 20, 2034). The NRC staff also finds that the defense-in-depth is maintained under the alternative through the performance of periodic bare metal visual examinations for evidence of head penetration leakage and through the enhanced online leakage detection capability of ANO-1.

Therefore, the NRC staff finds that removal of the RADCAL instrument to permit volumetric or surface examination of Penetration No. 1 in accordance with 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-6 would result in a hardship without a compensating increase in the level of quality and safety as specified in 10 CFR 50.55a(z)(2), and the proposed alternative would provide reasonable assurance of structural integrity of the penetration and the RVCH.

CONCLUSION The NRC staff has determined that complying with the specified requirements described in the licensees request referenced above would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The proposed alternative provides reasonable assurance of structural integrity of the subject component.

The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(2).

The NRC staff authorizes the use of proposed alternative ANO1-ISI-035 at ANO-1 to the end of the current, renewed ANO-1 operating license (May 20, 2034).

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 Contributors: M. Burton, NRR J. Collins, NRR Date: May 24, 2023 Digitally signed by Jennifer L. Jennifer L. Dixon-Herrity Date: 2023.05.24 Dixon-Herrity 13:27:20 -04'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation cc: Listserv

ML23142A106 *by email OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DNRL/NPHP/BC* NRR/DORL/LPL4/BC*

NAME TWengert PBlechman MMitchell JDixon-Herrity DATE 5/19/2023 5/24/2023 4/20/2023 5/24/2023