ML22118A686

NRR E-mail Capture - Davis-Besse Nuclear Power Station, Unit No. 1 - Request for Additional Information Regarding Alternative to Extend the Steam Generator Weld Inspection Interval

ML22118A686
Person / Time
Site:	Davis Besse
Issue date:	04/28/2022
From:	Blake Purnell Plant Licensing Branch III
To:	Lashley P FirstEnergy Nuclear Operating Co
References
Download: ML22118A686 (5)

Text

From: Purnell, Blake Sent: Thursday, April 28, 2022 1:38 PM To: Lashley, Phil H (EH)

Cc: Salgado, Nancy; kjnevins@energyharbor.com; kmnesser@energyharbor.com

Subject:

Davis-Besse Nuclear Power Station, Unit No. 1 - Request for Additional Information Regarding Alternative to Extend the Steam Generator Weld Inspection Interval Attachments: Davis-Besse 2nd RAI for SG Weld Inspection Alternative.pdf Mr. Lashley, By application dated September 13, 2021 (ADAMS Accession No. ML21256A119), as supplemented by letter dated January 27, 2022 (ADAMS Accession No. ML22027A770), Energy Harbor Nuclear Corp. (the licensee) submitted a request for a proposed alternative to certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, Codes and standards, for Davis-Besse Nuclear Power Station, Unit No. 1. Specifically, in accordance with 10 CFR 50.55a(z)(1), the application requests U.S. Nuclear Regulatory Commission (NRC) approval to increase the examination interval for the steam generator welds and nozzle inner radii from 10 years to 30 years. These examination requirements are specified in tables IWB-2500-1 and IWC-2500-1 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, as incorporated by reference in 10 CFR 50.55a.

The NRC staff is currently reviewing the application, as supplemented, and has determined that additional information is needed to complete this review. The NRC staff discussed the additional information needed with the licensee on April 27, 2022. During that discussion, the licensee stated that it could respond within 60 days. The NRC staff requests that a response to the attached request for additional information be provided within 60 days of the date of this email.

Please contact me if you have any questions.

Sincerely, Blake Purnell, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Docket No. 50-346 EPIDs L-2021-LLR-0067 OFFICE NRR/DORL/LPL3/PM NRR/DNRL/DD NRR/DNRL/D NRR/DORL/LPL3/BC NAME BPurnell BThomson BSmith NSalgado (JWiebe for)

DATE 4/28/22 3/16/22 3/16/22 3/16/22

Hearing Identifier: NRR_DRMA Email Number: 1613 Mail Envelope Properties (BLAPR09MB64971C493B2162D817C5D434E6FD9)

Subject:

Davis-Besse Nuclear Power Station, Unit No. 1 - Request for Additional Information Regarding Alternative to Extend the Steam Generator Weld Inspection Interval Sent Date: 4/28/2022 1:38:05 PM Received Date: 4/28/2022 1:38:00 PM From: Purnell, Blake Created By: Blake.Purnell@nrc.gov Recipients:

"Salgado, Nancy" <Nancy.Salgado@nrc.gov>

Tracking Status: None "kjnevins@energyharbor.com" <kjnevins@energyharbor.com>

Tracking Status: None "kmnesser@energyharbor.com" <kmnesser@energyharbor.com>

Tracking Status: None "Lashley, Phil H (EH)" <phlashley@energyharbor.com>

Tracking Status: None Post Office: BLAPR09MB6497.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1869 4/28/2022 1:38:00 PM Davis-Besse 2nd RAI for SG Weld Inspection Alternative.pdf 112574 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

REQUEST FOR ADDITIONAL INFORMATION PROPOSED ALTERNATE FOR EXAMINATION OF STEAM GENERATOR WELDS ENERGY HARBOR NUCLEAR GENERATION LLC ENERGY HARBOR NUCLEAR CORP.

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 By application dated September 13, 2021 (ADAMS Accession No. ML21256A119), as supplemented by letter dated January 27, 2022 (ADAMS Accession No. ML22027A770),

Energy Harbor Nuclear Corp. (the licensee) submitted a request for a proposed alternative to certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, Codes and standards, for Davis-Besse Nuclear Power Station, Unit No. 1. Specifically, in accordance with 10 CFR 50.55a(z)(1), the application requests U.S. Nuclear Regulatory Commission (NRC) approval to increase the examination interval for the steam generator (SG) welds and nozzle inner radii (Item Nos. B2.40, C1.30, C2.21, and C2.22) from 10 years to 30 years. These examination requirements are specified in Tables IWB-2500-1 and IWC-2500-1 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI, as incorporated by reference in 10 CFR 50.55a.

The proposed alternative is based on the methodology described in the Electric Power Research Institute (EPRI) Report No. 3002015906, Technical Bases for Inspection Requirements for PWR [Pressurized-Water Reactor] 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 (ADAMS Accession No. ML20225A141), and the EPRI Report No. 3002014590, Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, April 2019 (ADAMS Accession No. ML19347B107) (collectively, the EPRI Reports).

The licensees January 27, 2022, supplement was in response to an NRC staff request for additional information (RAI) issued on November 16, 2021 (ADAMS Accession No. ML21321A379). The NRC staff is reviewing the application, as supplemented, and has determined that additional information regarding the licensees response to RAI-12 is required to complete the review.

Regulatory Basis The regulations in 10 CFR 50.55a(g)(4)(ii) requires, in part, that inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals (i.e., after the initial 10-year interval) must comply with the latest edition and addenda of the ASME Code (or the optional ASME Code Cases) incorporated by reference in 10 CFR 50.55a(a) 18 months before the start of the 120-month inspection interval subject to the conditions listed in 10 CFR 50.55a(b). In accordance with 10 CFR 50.55a(z)(1), the NRC staff may authorize an alternative to an ASME Code,Section XI requirement established through 10 CFR 50.55a(g)(4)(ii) if the licensee demonstrates that the proposed alternative provides an acceptable level of quality and safety.

Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256), provides general guidance concerning analysis of the risk associated with proposed changes in plant design and operation, including the general principles of risk-informed decision-making: meeting current regulations, consistency with the defense-in-depth philosophy, maintaining safety margins, any increase in risk is small, and performance monitoring.

RAI-1

Issue The licensee referenced probabilistic and deterministic analyses (the EPRI Reports) to estimate potential fatigue growth in the subject weld and nozzle components. The licensee presented plant-specific information to demonstrate that the referenced EPRI analyses would bound the subject components, including high-level results from previous examination of the subject components. The licensee also provided limited discussion of performance monitoring, primarily focused on justifying application of the analyses to the proposed examination interval extension for the subject components (e.g., that leakage would be detected).

Leveraging probabilistic fracture mechanics (PFM) to define the basis for risk-informing inspection requirements requires knowledge of both the current and future behavior of the material degradation and the associated uncertainties applicable to the subject components.

Confidence in the results of these analyses hinges on the assurance that the PFM model adequately represents, and will continue to represent, the degradation behavior in the subject components. The NRC staff has determined that, when considering extended examination intervals, adequate performance monitoring through inspections is needed to ensure that the PFM model continues to predict the material behavior and that emergent degradation is discovered and dispositioned in a timely fashion.

The licensee discusses the system leakage test as a technical basis for the proposed alternative. However, the NRC staff notes that the visual examinations preformed during system leakage tests may not provide sufficient information to ensure that the PFM model continues to predict the material behavior and that emergent degradation is discovered and dispositioned in a timely fashion. Specifically, visual examinations may not directly detect pertinent integrity conditions (e.g., 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 components.

Request Describe the performance monitoring that will be implemented with this proposed alternative to ensure that the PFM model adequately represents, and will continue to represent, the degradation behavior in the subject components for the requested duration of the proposed alternative. Justify that this performance monitoring will meet this objective and address the concerns discussed above. Explain how this performance monitoring will provide, over the extended examination interval, (1) direct evidence of the presence and extent of degradation, (2) validation and confirmation of the continued adequacy of the PFM model; and (3) timely detection of novel or unexpected degradation. Describe any actions that will be taken if issues

are identified through this performance monitoring to ensure that the integrity of the components is adequately maintained.