ML25007A234

Limited Scope Aging Management Audit Report Regarding the Subsequent License Renewal Application Review

ML25007A234
Person / Time
Site:	Summer
Issue date:	01/17/2025
From:	Bill Rogers NRC/NRR/DNRL/NLRP
To:	Carr E Dominion Energy South Carolina
Shared Package
ML25007A232	List: ML25007A234
References
Download: ML25007A234 (11)

Text

January 17, 2025 Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 - LIMITED SCOPE AGING MANAGEMENT AUDIT REPORT REGARDING THE SUBSEQUENT LICENSE RENEWAL APPLICATION REVIEW

Dear Eric S. Carr:

By letter dated August 17, 2023 (Agencywide Documents Access and Management System Accession No. ML23233A175), Dominion Energy South Carolina, Inc. (DESC, Dominion Energy), on behalf of itself and Santee Cooper, submitted an application for the subsequent license renewal of Renewed Facility Operating License No. NPF-12 for Virgil C. Summer Nuclear Station, Unit No. 1 (V.C. Summer) to the U.S. Nuclear Regulatory Commission (NRC or staff).

The NRC staff performed a limited scope aging management audit in person from August 12, 2024 -

August 14, 2024, and continued virtually until October 25, 2024, in accordance with the audit plan (ML24109A179), concerning Aging Management Review, Further Evaluation Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation. The audit report is enclosed.

If you have any questions on this matter, please contact Marieliz Johnson, Project Manager, for the safety review of the V.C. Summer Subsequent License Renewal Application at Marieliz.VeraAmadiz@nrc.gov.

Sincerely,

/RA/

Bill H. Rogers, Chief (Acting)

License Renewal Projects Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation Docket No. 50-395

Enclosure:

Aging Management Audit Report cc: w/encl.: Listserv

PKG: ML25007A232 (PROP): ML25007A233 (Public): ML25007A234

• via email OFFICE PM: NRR/DNRL/NLRP*

LA: NRR/DNRL

• BC: NRR/DNRL/NLRP*

NAME MJohnson KBratcher BRogers DATE 1/06/2025 1/08/2025 1/17/2025

Audit Report Virgil C. Summer Nuclear Station, Unit No. 1 Subsequent License Renewal Application Limited Scope Aging Management Audit August 12, 2024 - October 25, 2024 Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

2 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF NEW AND RENEWED LICENSES Docket Nos:

50-395 License No:

NPF-12 Licensee:

Dominion Energy South Carolina, Inc.

Facility:

Virgil C. Summer Nuclear Station, Unit 1 Audit Location:

Dominion Energy Office 5000 Dominion Blvd, Glen Allen, VA 23060 Dates:

August 12 - October 25, 2024 Approved By:

Ian Tseng, Chief Structural, Civil, Geotech Engineering B Branch; Division of Engineering and External Hazards Angie R. Buford, Chief Vessels and Internals Branch; Division of New and Renewed Licenses Scott Krepel, Chief Nuclear Methods & Fuel Analysis Branch; Division of Safety Systems Reviewers:

Andrew Prinaris Structural, Civil, Geotech Engineering B Branch; Division of Engineering and External Hazards David Dijamco Vessels and Internals Branch; Division of New and Renewed Licenses Joe Messina Nuclear Methods & Fuel Analysis Branch; Division of Safety Systems Bill Rogers License Renewal Projects Branch; Division of New and Renewed Licenses

3 Report for the Limited Scope Aging Management Audit Virgil C. Summer Nuclear Station, Unit 1 Subsequent License Renewal Application

1. Introduction By letter dated August 17, 2023 (Agencywide Documents Access and Management System Accession No. ML23233A175), Dominion Energy South Carolina, Inc. (DESC), on behalf of itself and Santee Cooper, submitted an application for subsequent license renewal of Renewed Facility Operating License No. NPF 12 for Virgil C. Summer Nuclear Station, Unit No. 1 (V.C. Summer or the applicant) to the U.S. Nuclear Regulatory Commission (NRC or staff).

In accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, the NRC will perform an limited scope aging management regulatory audit to gain a better understanding of (1) the applicants methodology to identify the systems, structures, and components (SSCs) to be included within the scope of subsequent license renewal and subject to an aging management review (AMR), and (2) the applicants aging management programs (AMPs), AMR items, time-limited aging analyses (TLAAs), and associated bases and documentation as applicable.

License renewal requirements are specified in 10 CFR Part 54. Guidance is provided in NUREG-2192, Revision 0, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), dated July 2017, and NUREG-2191, Revision 0, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, dated July 2017.

Following the initial Aging Management Audit Plan (ML23296A109) and applicant responses to breakout questions (ML24085A701) as related to Subsequent License Renewal Application Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation, the staff determined additional review was needed. Accordingly, the staff conducted a limited scope audit to review the applicants responses and supporting bases to additional questions contained in the April 19, 2024, document, Supplemental Audit Questions, (ML24109A178), which are discussed in Section 2 of this report, Audit Activities.

2. Audit Activities Topic. Subsequent license renewal application (SLRA) AMR Further Evaluation Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation.

Summary of Information in the Application. Following its initial review and Audit of SLRA Section 3.5.2.2.2.6 the staff conducted a limited scope on-site audit to further clarify applicants determination that during the subsequent period of extended operation: (a) the [primary shield wall] PSW will continue to satisfy its design criteria considering the long-term radiation effects and a plant-specific AMP or enhancements to an existing AMP is not required; (b) a separate analysis of the [secondary shield wall] SSW is not required; (c) reduction of strength and loss of mechanical properties due to irradiation will not impact the primary shield walls intended

4 function under design basis conditions; (d) the RV supports continue to be structurally stable (i.e., flaw tolerant) considering 80 years of radiation embrittlement effects on the supports; (e) no additional inspections or enhancements are required for aging management of the RV supports, and the current ASME Code,Section XI inspection requirements are sufficient; and (f) examination of VCSNS operating experience precludes the synergism of other aging effects with those associate with radiation.

Audit Activities. During its on-site audit, the staff interviewed applicant (Dominion and Westinghouse Electric Co. (Westinghouse)) specialists on radiation, structural engineering, and fracture mechanics that contributed to the development of SLRA Section 3.5.2.2.2.6. In addition, the staff reviewed additional documentation supporting SLRA Section 3.5.2.2.2.6. The table below lists the additional documents that were reviewed by the staff and were found relevant to the Further Evaluation of SLRA Section 3.5.2.2.2.6. The staff will document its review of this information in the safety evaluation (SE).

Relevant Documents Reviewed Document Title Revision /

Date N/A (Proprietary)

TRP 76 Limited Scope Audit Questions and Responses 05/16/2024 N/A (Proprietary)

WEC Response to VC Summer SLRA Supplement Limited Scope Audit Questions 08/09/2024 WCAP-18785-P (Proprietary) /

WCAP-18785-NP V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal:

Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports Revision 2 WCAP-18757-P (Proprietary)

Technical Justification for Eliminating Auxiliary Piping Rupture as the Structural Design Basis Using Leak Before Break Methodology for VC Summer Nuclear Station Unit 1 June 2022 Calculation DC 0311E-013 (SCEG / Areva /

Westinghouse) (Proprietary)

Reactor Building Primary Shield Wall - Anchor Assembly Reactor Vessel Support / VC Summer Unit 1 RV Support Loads Evaluation /

Westinghouse Letter: Reactor Vessel Support Loads - VC Summer Unit 1

Revision 5 /

Revision 2 /

03/23/2016 CGE-CA120-TM-SA-000001 (Proprietary)

VC Summer Nuclear Station Unit 1: Subsequent License Renewal:

Primary Shield Wall Concrete Assessment Revision 2 GAI Calculation Master V.C. Summer, Reactor Building Interior Concrete/ Primary Shield Wall, Design Review 12/04/1975 GAI Calculation 1.51 -

Extract (pages 1-46/2382)

(Proprietary)

Reactor Building Interior Concrete Primary Shield Wall - Westinghouse Reactor Vessel Support Loads 07/16/1973 GAI Calculation 1.51 -

Extract (page 33/2382)

(Proprietary) : Structural Steel Load Conditions & Materials and Codes Used Revision 0 GAI Calculation 1.51 -

Extract (page 32/2382)

(Proprietary) : Reinforced Concrete Load Conditions August 2021 GAI Calculation 1:52(Proprietary)

Reactor Building Interior Concrete Primary Shield Wall Cavity Liner Revision 0 GAI Calculation 1:53(Proprietary)

Reactor Building Interior Concrete Reactor Vessel Spt. Anchor Assembly (Aka 1-DC-309-SPT-35M6D-0-Q)

Revision 0 VCSNS-DBD-RB South Carolina Electric & Gas Company, Virgil C. Summer Nuclear Station, Design Basis Document, Reactor Building Revision 9

5 WCAP-13206 (Proprietary) Technical Justification for Eliminating Large Primary Loop Pipe Rapture as the Structural Design Basis for the Virgil C. Summer Nuclear Power Plant Revision 4 WCAP-18772 (Proprietary) Resolution of Virgil C. Summer Nuclear Station Time-Limited Aging Analyses for Subsequent License Renewal Revision 1 ETE-SLR-2023-3336 Subsequent License Renewal Project Aging Management Program Evaluation Report ASTM E Section XI, Subsection IWF V.C. Summer Nuclear Station Revision 0 ES-0439 Engineering Services Procedure - In Service Inspection Program Controls Revision 1 CGE-REAC-CN-AA-000001 (proprietary)

Virgil C. Summer Reactor Vessel, Vessel Support, and Bioshield Concrete Exposures for Subsequent License Renewal Revision 1 RCR C-00-1392 V.C. Summer Nuclear Station, Root Cause Investigation, A Hot Leg Nozzle Weld Cracks 02/15/2001 1-DC-309-SPT M6D Q Structural Engineering Calculation No. 1.53. Reactor Building Interior Concrete Area. Reactor Vessel Support Anchor Assembly/Anchor Assembly under RV Support - Primary Shield Wall Revision 0, 11/15/76 SP-625-044461-000 Specification for Special Steel for Reactor Building, Virgil C.

Summer Nuclear Station Unit 1 05/12/1975 AB 15517 (Proprietary)

Engineers Technical Work Record: Boric Acid Cleaning and Inspections per CER-00-1234 11/30/00 VCS-PTP-151.001 Preventive Test Procedure: Inspection for Boric Acid Corrosion (includes Unit 1 Reactor Building Area Inspection Report - Step 8.1.2, Reactor Vessel Lower Head, Incore Pit)

Revision 1 QSP-216 Quality Systems Procedure: Boric Acid Corrosion Inspection Revision 4 ES-0437 Engineering Services Procedure: Inspections for Maintenance Rule Structures (Part O.: Inaccessible Building Structures Associated with Reactor Vessel Nozzle Supports and RCS Piping Embedded LOCA Supports)

Revision 3 NCN 00-1603 Disposition 1 (Proprietary)

Engineers Technical Work Record: Cavity Seal Ring 01/17/2001 N/A Hilsdorf, H.K., Kropp, J., and Koch, H.J, The Effects of Nuclear Radiation on the Mechanical Properties of Concrete, American Concrete Institute SP 55-10, p223-234 1978 N/A Bruck et al., Structural Assessment of Radiation Damage in Light Water Power Reactor Concrete Biological Shield Walls, Nuclear Engineering and Design, Volume 350, pages 9-20 08/15/2019 DWG E-511-110 (Proprietary)

VCSNS Unit #1: Reactor Building Steel Framing In-Core Instrumentation Supports Revision 13 DWG 1MS-07-159 590063-001C (Proprietary)

RV Insulation General Arrangement for Westinghouse multi vessel plant

1. 2 (Diamond Power/Babcock and Wilcox)

Revision A DWG 1MS-07-129 (Proprietary) 157 PWR Inlet Nozzles Revision 7 DWG 1MS-07-1296 (Proprietary)

Reactor Vessel Assembly Outlet Nozzle Revision 1 01/02/1997)

DWG 1MS-15-163 (Proprietary)

Cavity Seal and Toggle System Assembly Revision 0 DWG E-922-004 (Proprietary)

VC Summer Reactor Building Primary Shield Wall Air Boots - Plans, Section and Detail Revision 2 During the Audit, the staff made the following observations for the primary shield wall concrete and associated areas of the reactor vessel steel supports:

6 Effects of Radiation on Steel and Reactor Cavity Cementitious Materials (Concrete, Grout)

The applicant indicated that approximately the initial four inches of the PSW were considered to be non-functional.

Westinghouse indicated that there was uncertainty and lack of clarity how the original Gilbert design proceeded into yielding the design Analysis of Record for the PSW reinforced concrete. In the absence of this knowledge Westinghouse made assumptions to generate its own Subsequent License Renewal Analysis to update the original demands to capacity ratios for the PSW concrete evaluation at 72 effective full-power years. Westinghouse then deemed their approach as proprietary. A lack of alignment and clarity of CGE-CA120-CN-SA-000001, Rev. 1, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment (W-AOR) with the Gilbert design Analysis of Record conveyed an uncertainty to the staff with respect to the SLRA-provided estimation of the demand to capacity ratios.

Westinghouse, based on this lack of knowledge as used and approximating approach, developed margins against component loss of intended function and demands for capacity.

Subsequently, the applicant provided information related to its Reactor Vessel (RV) head replacement and leak-before-break analyses that indicated information of reduced loads that potentially could have revised the demand to capacity ratios downwards, but the staff determined that this information was not adequate to support the lack of proposed aging management activities since these effects of reduced loading were not incorporated in the applicants analysis summarized in the SLRA.

Effects of Ventilation on the PSW Concrete The applicant discussed ventilation of the reactor cavity and its adequacy to reduce potential detrimental Gamma Heating effects on the PSW concrete.

Adverse Environments on Cementitious Materials in the Reactor Cavity To determine whether cementitious materials in the reactor cavity (concrete and grout) had been compromised in the past, the NRC staff performed a detailed review of the two below Operating Experience (OE)s:

a) Effects of air temperature on the PSW concrete - event-based on operating experience (2000): This OE is relevant to an elevated temperature in the reactor cavity due to operational failure of annulus ventilation. The applicant provided evaluations of the PSW concrete and its embedded steel which were exposed to 250 degrees Fahrenheit (°F) for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. Summarized calculations indicate that the concrete surface near the reactor vessel supports rose to 233°F during that time.

b) Effects of Boric Acid on Reactor Vessel Supports (accessible and inaccessible) and reinforcement - event-based on operating experience (1996): This OE is relevant to an event that lasted 18 months, in which a crack developed near a reactor coolant system component (vessel nozzle containing boric acid) on the

7 hot leg. Review of documents and discussion with the applicant indicated that the reactor vessel supports had been cleaned and evaluated.

Effects of Radiation of Steel Embedded in the PSW Concrete The applicant provided numerous documents for audit (e.g., WCAP-18785-P (Revision 2), GA DWG E-511-214) and discussions to put forth the position provided in various evaluations that the embedded steel did not require any management activities.

Discussion on the intended functions of the embedded steel (i.e., wide-flange sections within the PSW) indicated that the wide flanges provided lateral load support to the reactor vessel supports. The staff confirmed this in audited GAI AOR Calculation No.

1.53 and Calculation No. DC 0311E-013.

The applicant stated that the GAI AOR loads in proprietary Calculation No. 1.51 for the reactor vessel supports were calculated ((

))

and are without implementation of leak-before-break.

Proprietary Calculation No. 1.51 clarifies the applied dead, thermal, Westinghouse operational basis earthquake/design basis earthquake and accident loads, as well as the source of the Accident load for the reactor vessel supports. The applicant stated that the design calculation for the embedded wide-flange steels in proprietary Calculation No.

1.53 conservatively did not include composite action.

The audited Specification for Special steel for Reactor Building Virgil C. Summer Nuclear Station Unit 1 require that the embedded wide-flange steels, SP-625-044461-000, states the following:

a) All plates of ASTM A302 material shall be tested for internal discontinuities by the plate manufacturer using ultrasonic inspection. (Section 2:06.1) b) Welding of plates shall use either electrode specification E70xx or E80xx.

(Sections 2:05.8 and 2:05.9) c) Weldments of ASTM A302 material shall be post-weld stress relieved. (Section 2:07.1, #8) d) All welds shall have 100 percent visual inspection. (Section 2:08.1, #4) e) All full penetration weld butt welds shall be 100 percent inspected by ultrasonic testing. (Section 2:08.1, #7) f) Magnetic particle tests on demand (e.g., GAI DWG E-511-213, Reactor Building Reactor Vessel Support Anchor Plan Elevation 426-6, Revision 2, indicates that 20% magnetic particle testing on welds on random selected basis will be required as advised by the engineer), are in accordance with ASTM E-109 and depending on the weld size inspections were performed on the root pass, after each 1/4-inch deposit, and/or final surface

8 Fracture Toughness of Plate Components of RV Steel Supports not Embedded within the PSW Concrete In SLRA 3.5.2.2.2.6, in the section titled Fracture Mechanics Evaluation, the applicant stated that the fracture toughness used for the plate components is based on the ASME Code,Section XI, lower bound KIC fracture toughness value of 33.2 kilopounds per square inch square root inches (ksiinch) The applicant did not confirm in the audited fracture mechanics evaluation of the RV steel supports not embedded in concrete (WCAP-18785-P/NP) that the ASME Code,Section XI, lower bound KIC fracture toughness value of 33.2 ksiinch was demonstrated to be bounding for the plant-specific material of the plate components of RV steel supports. The NRC staff determined that this information needs to be provided on the docket.

Irradiation Exposure Estimates Uncertainties The applicant provided justification for the 20 percent uncertainty applied to the neutron fluence and gamma dose projections in the PSW and the 25 percent uncertainty applied to the dpa projections to the support structure (i.e., the support box plate, support box, support show, and support box plate bolt). The applicants justification aligned with the justification provided in the previous audit, but the NRC staff stated that the justification needs to be provided on the docket.

Conclusion. Based on the preceding information, the staff indicated that it had determined that there was not sufficient information in the SLRA, or provided in the audit, to conclude reasonable assurance and preclude necessitation of aging management activities for the concrete, steel, and grout comprising the reactor vessel supports.

The applicant indicated that they would provide a voluntary SLRA supplement for the necessary information to address issues identified and communicated during the audit. The staff will document its review of this information and resolution of the issues in the safety evaluation report.

3. Applicant Personnel Contacted During the Audit Name Affiliation Keith Miller Dominion Energy Daniel Madden Dominion Energy Michael Guthrie Dominion Energy Julian Hamilton Dominion Energy Cindi Henderson Dominion Energy David Clohecy Dominion Energy Allen Hiser Dominion Energy Charles A. Tomes Dominion Energy Paul Phelps Dominion Energy Xi Liu Westinghouse Anees A. Udyawar Westinghouse Daniel Carlin Westinghouse Todd Baker Westinghouse Ryan Griffin Westinghouse

9

4. Exit Meeting An exit meeting was held with the applicant on October 25, 2024, to discuss the results of the regulatory audit. The applicant submitted an SLRA supplement (ML24302A144) dated October 24, 2024, addressing the topics discussed in the audit. The SLRA supplement will be evaluated by the staff in the SLRA SE.