ML22300A220

(VCSNS) Unit 1, Alternative Request Rr-4-26 Application for Alternative Request - Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination) RAI

ML22300A220
Person / Time
Site:	Summer
Issue date:	10/27/2022
From:	James Holloway Dominion Energy South Carolina
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
22-312
Download: ML22300A220 (1)

Text

Dominion Energy South Ca rolina, Inc.

5000 Dominion Boulevard, Gle n Allen, VA 23060 Domin ionEnergy.com October 27, 2022 a~

Dominion Energy*

Attn: Document Control Desk Serial No.: 22-312 U. S. Nuclear Regulatory Commission NRA/YG: RO Washington, DC 20555-0001 Docket No.: 50-395 License No.: NPF-12 DOMINION ENERGY SOUTH CAROLINA (DESC}

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS} UNIT 1 ALTERNATIVE REQUEST RR-4-26 "APPLICATION FOR ALTERNATIVE REQUEST

- EXTENSION OF STEAM GENERATOR PRIMARY INLET NOZZLE DISSIMILAR METAL WELD INSPECTION INTERVAL (VOLUMETRIC EXAMINATION}"

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

By letter dated March 10, 2022 (Agencywide Document Access and Management System Package Accession No. ML22069B117), Dominion Energy South Carolina, Inc. (DESC),

submitted an alternative request to the volumetric examination In-Service Inspection (ISi) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI for the Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Class 1 PWR pressure retaining dissimilar metal (DM) piping and vessel nozzle butt welds.

Specifically, DESC proposed a one-time extension of the VCSNS steam generator (SG) primary inlet nozzle dissimilar metal weld (DMW) inspection interval from 5 years to nominal 9 years for the volumetric examination. No changes were requested to the visual examination requirements applicable to the SG inlet nozzle DMW. DESC was seeking to defer performance of the volumetric examination from the current scheduled date in the spring of 2023 to the fall of 2027.

In an email dated September 19, 2022, from Ed Miller, NRG Senior Project Manager, to Yan Gao of Dominion Energy, the Nuclear Regulatory Commission (NRG) staff requested additional information to facilitate their review of the subject alternative request. The NRC's request for additional information (RAI) and the DESC responses are provided in the attachment to this letter. contains the items provided in the RAI, the corresponding responses, and the applicable references.

Serial No.22-312 Docket No. 50-395 Page 2 of 3 Should you have any questions, please contact Yan Gao at (804) 273-2768.

Respectfully, Jam1::way Vice President - Nuclear Engineering and Fleet Support Commitments made in this letter: None.

Attachment:

1. Response to NRG Request for Additional Information

Serial No.22-312 Docket No. 50-395 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector V.C. Summer Nuclear Station Ms. Anuradha Nair-Gimmi Bureau of Environmental Health Services South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461

Serial No.22-312 Docket No. 50-395 Attachment 1 Response to NRC Request for Additional Information Virgil C. Summer Nuclear Station {VCSNS) Unit 1 Dominion Energy South Carolina, Inc. {DESC)

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 1 of 7 TABLE OF CONTENTS

1.0 BACKGROUND

.. ... ....................................... ... ................................................................ 2 2.0 NRC RAIS ..... ************************* ..................... ... .................................... **** ***** ........... .......... 2 2.1 RAI Description ............................................................................................................ 2 2.2 RAI Responses .. ........................................ ... .................................. ............................. 3

3.0 REFERENCES

.. ........................... .... .... ...... ............................... ...................................... 7

Serial No.22-312 Docket Nos. 50-395 Attachment 1: Page 2 of 7 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - APPLICATION FOR ALTERNATIVE REQUEST-EXTENSION OF STEAM GENERATOR PRIMARY INLET NOZZLE DISSIMILAR METAL WELD INSPECTION INTERVAL

{VOLUMETRIC EXAMINATION)

DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1

1.0 BACKGROUND

By letter dated March 10, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML220698117) [3.4], Dominion Energy South Carolina, (DESC), submitted an alternative request (AR) for the Virgil C. Summer Nuclear Station, Unit 1, (VCSNS), for a one-time extension of the VCSNS steam generator (SG) primary inlet nozzle dissimilar metal weld (DMW) inspection interval from 5 years to nominal 9 years for the volumetric examination. No changes were requested to the visual examination requirements applicable to the SG inlet nozzle DMW. DESC was seeking to defer performance of the volumetric examination from the current scheduled date in the spring of 2023 to the fall of 2027. The U.S. Nuclear Regulatory Commission (NRG) staff has reviewed the AR and generated a request for additional information (RAI) [3.5] in order to complete the review. Section 2 below provides the RAI items and the corresponding responses.

2.0 NRC RAls 2.1 RAI Description By letter to U.S. Nuclear Regulatory Commission (NRG) dated March 10. 2022 (Agencywide Documents Access and Management System Accession No. ML22069B117). Dominion Energy South Carolina Inc. (the licensee) proposed an alternative to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-770-5. "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Filler Material With or Without Application of Listed Mitigation Activities." The licensee submitted Proposed alternative RR-4-26. for an extension of the steam generator primary inlet nozzle dissimilar metal weld inspection interval from 7 years to 9 years for the volumetric examination.

The proposed alternative uses a Factor of Improvement (FOi) for the crack growth rate of alloy 521152, as compared to alloy 821182 weld metals. To date. the NRG has approved the use of an FOi of 10 for alloy 152 weld materials. The NRG is currently working with the Primary Water Stress Corrosion Cracking (PWSCC) Crack Growth Rate Expert Panel, which is evaluating the crack growth PWSCC data for Alloys 52 and 152, including several data sets not included in the letter dated March 10. 2022. The NRG

• RAI texts are italic and underlined

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 3 of 7 does not yet have sufficient information to support the use of a FOi of 29. The NRG expects, however, new information on the appropriate FOi will likely be published in 2023.

The NRG has determined that the following additional information is necessary to complete its review and make a regulatory decision on RR-4-26.

1) Beyond the FOi approach, what additional defense-in-depth measures would be taken by the licensee to support a limited extension of examination interval until the expert panel completes its activities.

2) Please provide additional justification to support an FOi of 29 or provide an alternative duration for the proposed alternative that would not require an FOi of 29.

2.2 RAI Responses Response to 2.1.1J

• Bare metal visual (BMV) examination of the VCSNS SG primary inlet nozzle dissimilar metal welds (DMWs) will continue to be performed every refueling outage in accordance with ASME Code Case N-722-1 [3.1] and ASME Code Case N-770-5

[3.2]. A BMV examination is scheduled to be performed on these DMWs in the upcoming Spring 2023 outage. In addition, these DMWs are subject to VT-2 examination at the end of every refueling outage prior to plant start-up as part of the RCS Mode 3 (ascending) System Leakage Test. In the unlikely event that leakage was to occur due to a through-wall flaw in the DMWs, evidence of RCS leakage would be detected by these ASME Code visual examinations. For the entirety of the VCSNS replacement SG service history, there has been no leakage detected in any of the DMWs.

• The VCSNS Boric Acid Corrosion Control (BACC) Program uses visual examinations to identify locations where RCS leakage smaller than allowable Technical Specification (TS) limits could cause degradation of the RCS pressure boundary due to boric acid corrosion. In the unlikely event that leakage was to occur due to a through-wall flaw in the VCSNS Steam Generator primary inlet nozzle DMWs, evidence of RCS leakage would be detected by these ASME Code visual examinations.

• Containment radiation monitors and containment sump level indication are required to be available during plant operation to detect substantial RCS leakage and/or increases in radiation levels within containment. In accordance with WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors [3.3], VCSNS monitors and trends absolute unidentified leak rate, deviation from the baseline mean, and total integrated unidentified leakage. Action levels are established to address leakage well before approaching the VCSNS TS limits defined in TS 3.4.6.2, REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE. These detection methods and TS requirements provide early detection of leakage from the RCS and minimize the consequences associated with RCS leakage.

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 4 of 7 Response to 2.1.2)

Using the same inputs and methodology as discussed in Reference [3.4], two (2) alternative durations for the proposed alternative that would not require a Factor of Improvement (FOi) of29 have been analyzed. An inspection interval of6 calendar years requires an FOi of 15 (deferral to Fall 2024), and an inspection interval of 7.5 calendar years requires an FOi of 23 (deferral to Spring 2026). These inspection intervals are based on the last volumetric examinations of the VCSNS SG primary inlet nozzle DMWs which were performed during the Fall 2018 refueling outage. Figures B-1 and B-2 show the PWSCC growth curves for an FOi of 15 (axial and circumferential flaws, respectively).

Figures B-3 and B-4 show the PWSCC growth curves for an FOi of 23 (axial and circumferential flaws, respectively).

In addition, the analyses performed for the 6 and the 7.5 year inspection intervals conservatively assumed that immediately following the last volumetric examinations (Fall 2018), crack initiation had already occurred in the Alloy 152 inlay and crack growth began immediately. The VCSNS SG primary inlet nozzle DMWs were examined in Fall 2018 using inside surface phased array ultrasonic volumetric examination and automated eddy current surface examination methods with no indications. To date, there have been no occurrences of PWSCC crack initiations in Alloy 152 weld materials.

0.8 0.7 0.6 Alio-wable Flaw Size Reached in

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o 2 3 4 Effecth-e F ull Po,..er Years (EFPY}

5 6 7 Figure B-1: Axial Flaw Growth Chart for SG Inlet Nozzle (T = 619.4°F) Aspect Ratio (AR)=2 For Alloy 152 PWSCC growth rate a FOi of15 over the Alloy 182 crack growth rate is used.

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 5 of 7 0.8 0.7

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Maximum Allowable End-of-fa.ilu.ation Period Flaw Size 0.6 Allowable Flaw Size Reached in 6.05EFPY 0.2 0.1 ob====================-- - - ~-----------

o 2 4 Effecfu-e Full Po..-er Years (EFPY) 6 7 Figure B-2: Circumferential Flaw Growth Chart for SG Inlet Nozzle (T=619.4°F) Aspect Ratio (AR)=IO For Alloy 152 PWSCC gruwth rate a FOI of15 over the Alloy 181 crack gruwth rate is used o.s

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Figure B-3: Axial Flaw Growth Chart for SG Inlet Nozzle (T=619.4°F) Aspect Ratio (AR)=2 For Alloy 152 PWSCC growth rate a FOI of13 owr the Alloy 181 crack growth rate is used

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 6 of 7 0.8 Maximum Allowable End-of-fa'3htilion Period Flaw Size 0.7 0.6 Allowable Flaw Size Reached in 7.48 EFPY*

0.2 0.1 ot==::~=====:::::===========~=------------_J 0 2 4 5 Ii 8 EffediTe Full Pon-er Ye:ars (EFP1.1 Figure B-4: Circum.feren tiaJ Flan- Grom .b Cluirt for SG Inlet Nozzle (T=<ll9.4°F) Aspect Ratio (AR)=IO For Alloy 152 PWSCC g..ov.-fl, rate a FOI of23 over the Alloy 182 crack growth rate is ri=i..

• A nll>.e of 7.48 EFPYis b ~ o:r I ~ c,pllltmgpcnnr, whxh isc~ m-e smce Ibo! plmt ,.,;n be in emu.~* fm , 8'!Dl morubsonrm. com5eof75Cll!IMW".fllll.

Serial No.22-312 Docket No. 50-395 Attachment 1: Page 7 of 7

3.0 REFERENCES

3.1 ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 MaterialsSection XI, Division 1 3.2 ASME Code Case N-770-5, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1 3.3 ML070310082, PWROG report, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors, WCAP-16465-NP Revision 0, September 2006 3.4 ML220698117, Alternative Request (AR) for the Virgil C. Summer Nuclear Station (VCSNS), Unit 1, Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval, March 10, 2022 3.5 Email from Ed Miller (NRC) to Yan Gao (Dominion) regarding RAI for VCSNS AR-4-26, September 19, 2022